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Sample records for advanced passive reactor

  1. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  2. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Science.gov (United States)

    2013-07-10

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised......

  3. Overview of human factors issues associated with advanced passive reactors

    International Nuclear Information System (INIS)

    Nuclear power plant operators in current plants do just that, they operate the plant. They monitor the state of the process, observe trends, react to system changes and take action to maintain the plant in a safe condition. In order to accomplish these activities, they (1) rely on information available to them from the various instruments and displays in the control room, (2) use hard copy procedures to direct their actions and (3) perform their control actions through the various switches, keyboards and knobs available. Proposals from vendors and observations of activities in other countries indicate that the control room for advanced passive reactors is likely to be quite different from control rooms one would seen in current power plants. These differences will go beyond the immediate appearances, e.g., smaller, more cockpit-like, to issues of automation, function-allocation, the changing role of the operators and their training and qualifications, and software verification and validation. This research paper discusses some of these planned efforts and how they relate to the human factors issues and regulatory user needs

  4. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Science.gov (United States)

    2012-10-12

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does... Section 19.3, ``Regulatory Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light......

  5. Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De, Cheng, E-mail: 0100209064@sjtu.edu.cn; Zhen-Qiang, Yao, E-mail: zqyaosjtu@gmail.com; Ya-bo, Xue; Hong, Shen

    2014-10-15

    Highlights: • An artificial accelerogram of the specified SSE is generated. • A dynamic FE model of the RCP in AP1000 (with gyroscopic and FSI effects) is developed. • The displacement, force, moment and stress in the RCP during the earthquake are summarized. - Abstract: The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed.

  6. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    International Nuclear Information System (INIS)

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

  7. Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

    International Nuclear Information System (INIS)

    Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

  8. Modern passive safety system for the advanced fast reactors with sodium cooling

    International Nuclear Information System (INIS)

    In fast reactors with sodium coolant it is possible to avoid serious damages of a core even at the heaviest scripts of development of accidents if to provide influence on reactance with the help of various type of Passive Safety System (PSS). With input of the PSS the reactor gets an additional negative feedback on the reactance, working at an output of the basic operational parameters (temperature, the coolant flow rate, power) for maximum permissible sizes. The scientific-technical and patent sources analysis has shown, that now it is already offered more than two hundred the various devices, capable to carry out functions of the fast reactors PSS. Comparison of various types of PSS is carried out under 9 generalized characteristics including: passivity, thresholdness, generation of efforts, inertia, multi-channels, stability to operational factors, refusal safety, simplicity and presentation, development conditions. For quantitative comparison of the device ''the perfection degree'' (K≤1) was defined as average size under 9 generalized characteristics. From the considered types of fast reactors PSS the most perfect now are fusible Lyophobic devices, basically meeting the requirements on all characteristics. Results of Lyophobic Passive Safety System development for the advanced fast reactors with sodium cooling are considered. Serviceability of the offered designs is proved experimentally at various operation temperatures on breadboard models sylphon devices and devices of type the sylphon-container with various lyophobic liquids: alloy Wuds (Tmt=80,0 deg C), an alloy lead-bismuth (Tmt=123,5 deg C), cadmium (Ttm=320,0 deg C), aluminium (Tym=660,0 deg C), developed Lyophobic Fusible Passive Safety System on excess of temperature are of interest for nuclear power installations of various type, first of all, as passive devices scram reactor and protection of the process equipment. (author)

  9. Requirements for Prognostic Health Management of Passive Components in Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. aSMRs are conceived for applications in remote locations and for diverse missions that include providing process or district heating, water desalination, and hydrogen production. Several challenges exist with respect to cost-effective operations and maintenance (O&M) of aSMRs, including the impacts of aggressive operating environments and modularity, and limiting these costs and staffing needs will be essential to ensuring the economic feasibility of aSMR deployment. In this regard, prognostic health management (PHM) systems have the potential to play a vital role in supporting the deployment of aSMR systems. This paper identifies requirements and technical gaps associated with implementation of PHM systems for passive aSMR components.

  10. Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project

    Energy Technology Data Exchange (ETDEWEB)

    Brian G. Williams; Jim C. P. Liou; Hiral Kadakia; Bill Phoenix; Richard R. Schultz

    2007-02-27

    This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

  11. Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R and D Project

    International Nuclear Information System (INIS)

    This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems

  12. Interim results of the study of control room crew staffing for advanced passive reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, B.P.; Sebok, A.; Haugset, K. [OECD Halden Reactor Project (Norway)

    1996-03-01

    Differences in the ways in which vendors expect the operations staff to interact with advanced passive plants by vendors have led to a need for reconsideration of the minimum shift staffing requirements of licensed Reactor Operators and Senior Reactor Operators contained in current federal regulations (i.e., 10 CFR 50.54(m)). A research project is being carried out to evaluate the impact(s) of advanced passive plant design and staffing of control room crews on operator and team performance. The purpose of the project is to contribute to the understanding of potential safety issues and provide data to support the development of design review guidance. Two factors are being evaluated across a range of plant operating conditions: control room crew staffing; and characteristics of the operating facility itself, whether it employs conventional or advanced, passive features. This paper presents the results of the first phase of the study conducted at the Loviisa nuclear power station earlier this year. Loviisa served as the conventional plant in this study. Data collection from four crews were collected from a series of design basis scenarios, each crew serving in either a normal or minimum staffing configuration. Results of data analyses show that crews participating in the minimum shift staffing configuration experienced significantly higher workload, had lower situation awareness, demonstrated significantly less effective team performance, and performed more poorly as a crew than the crews participating in the normal shift staffing configuration. The baseline data on crew configurations from the conventional plant setting will be compared with similar data to be collected from the advanced plant setting, and a report prepared providing the results of the entire study.

  13. Typical technology of mechanics on Gen-III passive NPPs and Gen-IV advanced supercritical light water reactors

    International Nuclear Information System (INIS)

    Full text: Technical requirements for Gen-III advanced nuclear power plants, which take passive reactors as the main body, were originally brought forward in American 'Advanced Light Water Reactor Utility Requirement Document' (ALWR-URD) in early 1990's. The primary characteristic of passive nuclear power plant is large amount of simplification to the original active safety systems, replacing or supplementing them with passive safety systems, which enhances safety and economy. However, the replacement of active safety systems by passive safety systems also brings about some mechanics that compel attention, typically, such as load-carrying capability evaluation for steel containment, in-vessel retention (IVR) of molten core debris, seismic design without OBE, thermo-hydraulic issues concerning with coupling between two-phase fluid and solid, etc. At the beginning of this century, six typical Gen-IV advanced reactor types (Sodium Cooled Fast Reactor, Supercritical Water-Cooled Reactor, etc.) were put forward. Among these types of reactors, Supercritical Water-Cooled Reactor adopts supercritical water as coolant and operates above the thermodynamic critical point of water by increasing temperature and pressure of the coolant, which makes the plant economic and efficient. However, this type of reactor also brings about some mechanical difficulties (e.g. pressure fluctuation caused by the supercritical fluid in the core, creep of materials working at high temperature, etc.) for the design of facility and components. In this paper, the issues mentioned above are outlined for further consideration. (author)

  14. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  15. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and

  16. ACP100 (China National Nuclear Corporation, China) [Passive Safety Systems in Advanced Small Modular Reactors

    International Nuclear Information System (INIS)

    The ACP100 is being developed by China National Nuclear Corporation (CNNC). It is an integral pressurized water reactor (iPWR) with a rated power of 100 MW(e). The reactor is proposed to be utilized for electricity generation, heat or desalination. A plant utilizing the design will have a flexible configuration, with between one and eight modules. A number of passive systems have been incorporated in ACP100. Some of them are described

  17. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    Energy Technology Data Exchange (ETDEWEB)

    Winters, J. W.; Corletti, M. M.; Hayashi, Y.

    2003-11-12

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

  18. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    International Nuclear Information System (INIS)

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States

  19. Experimental and analytical studies of passive shutdown heat removal from advanced LMRs [liquid metal reactors

    International Nuclear Information System (INIS)

    A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) has investigated the heat transfer performance of the GE/PRISM passive design. This initial series of experiments simulates the air-side geometry of the PRISM Radiant Reactor Vessel Auxiliary Cooling System (RVACS). The NSTF operates in either a uniform heat flux mode and a uniform temperature mode at the air/guard vessel interface. Analysis of the RVACS performance data indicates excellent agreement with pretest analytical predictions. Correlation analysis presents the heat transfer data in a form suitable for use in LMR design and verification of analytical studies

  20. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  1. Fundamental study on thermo-hydraulic phenomena concerning passive safety of advanced marine reactor

    International Nuclear Information System (INIS)

    The objective of this study is to investigate the thermo-hydraulic behavior of a fluid region confined in a rectangular parallelepiped cavity equipped with a heater and a cooler. The motivation of this study is to clarify a thermal buffer effect for an innovative marine nuclear reactor to realize passive safety. In the present study, experiments were carried out with conditions of laminar convection. Temperature and flow behavior was visualized by the liquid-crystal suspension method, by which the temperature distribution in liquid can be observed as a colored map. Thermal plumes from the heater and the cooler, global natural circulation in the cavity and thermal stratification were observed as elements of the complicated phenomena. Using a code which solves the Navier-Stokes and energy equations, numerical simulations under steady and unsteady condition were carried out to predict the experimental results for two-dimensional, laminar situations, and a good agreement was obtained. (author)

  2. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    International Nuclear Information System (INIS)

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  3. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  4. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  5. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shripad T. Revankar; Seungmin Oh

    2003-09-30

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  6. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  7. Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

  8. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  9. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to: (1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, (2) assess the RELAP5 and TRACE computer code against the experimental data, and (3) develop mathematical model and heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

  10. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  11. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

  12. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

  13. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  14. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  15. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  16. Passive containment cooling for an advanced PWR

    International Nuclear Information System (INIS)

    The AP600 is a 600-MW(electric) pressurized water reactor that is currently being developed by Westinghouse and its subcontractors. The AP600 program is being sponsored by the US Department of Energy (DOE) in conjunction with DOE and Electric Power Research Institute advanced light water reactor programs. The AP600 employs safety features that, when actuated, use natural phenomena and stored energy (gravity, natural circulation, compressed gas) to accomplish all required safety functions. This safety approach results in both improved safety and a significant simplification in the overall plant design since no safety-grade ac power or support systems are required. Also, significant reductions in plant complexity, capital cost, and construction schedule can be achieved. One of the key safety systems in the AP600 passive safety approach is the passive containment cooling system (PCCS). The PCCS provides the safety-grade ultimate heat sink for the removal of reactor-sensible heat and core decay heat following any design-basis event. Analytical models of the PCCS have been developed and transient and accident evaluations have been performed to demonstrate the heat removal capability to the PCCS. These analyses indicate that AP600 postaccident containment response is similar to that achieved with active containment heat removal systems. Also in conjunction with analyses, a test program is under way to demonstrate and verify the heat removal capability of the PCCS design concept

  17. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    International Nuclear Information System (INIS)

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

  18. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  19. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  20. U. S. Utility Leadership in Requirements For Passive Reactors

    International Nuclear Information System (INIS)

    Utility leadership from both U.S. utilities and international utilities, is a key element in the U. S. Advanced Light Water Reactor Program. International utilities have played a very import Design reviews by the utilities participating in the ALRR Program will ensure that all of the utility requirements are met while design work is being carried out. Our mission is to achieve NRC certification of designs that reflect the needs of the utilities and we believe that this will play an important role in the resurgence of nuclear plant construction in the United States. As stated in the Nuclear Power Oversight Committee's Strategic Plan For Building New Nuclear Power Plants : 'The extensive operating experience with today's light water reactors (LWRs), and the promise shown in recent technical developments, leads the industry to the conclusion that the next nuclear plants ordered in the United States will be advanced light water reactors (A LWRs). Two types are under development : units of large output (1300 MW) called 'evolutionary' A LWRs and units of mid-size output (600 MW) called 'Passive' A LWRs. The term 'passive' refers to the safety features which depend more on natural processes such as gravity and buoyancy than on powered equipment such as pumps

  1. Engineering design of advanced marine reactor MRX

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    JAERI has studied the design of an advanced marine reactor (named as MRX), which meets requirements of the enhancement of economy and reliability, by reflecting results and knowledge obtained from the development of N.S. Mutsu. The MRX with a power of 100 MWt is intended to be used for ship propulsion such as an ice-breaker, container cargo ship and so on. After completion of the conceptual design, the engineering design was performed in four year plan from FY 1993 to 1996. (1) Compactness, light-weightiness and simplicity of the reactor system are realized by adopting an integral-type PWR, i.e. by installing the steam generator, the pressurizer, and the control rod drive mechanism (CRDM) inside the pressure vessel. Because of elimination of the primary coolant circulation pipes in the MRX, possibility of large-scale pipe break accidents can be eliminated. This contributes to improve the safety of the reactor system and to simplify the engineered safety systems. (2) The in-vessel type CRDM contributes not only to eliminate possibilities of rod ejection accidents, but also to make the reactor system compact. (3) The concept of water-filled containment where the reactor pressure vessel is immersed in the water is adopted. It can be of use for emergency core cooling system which maintains core flooding passively in case of a loss-of-coolant accident. The water-filled containment system also contributes essentially light-weightness of the reactor system since the water inside containment acts as a radiation shield and in consequence the secondary radiation shield can be eliminated. (4) Adoption of passive decay heat removal systems has contributed in a greater deal to simplification of the engineered safety systems and to enhancement of reliability of the systems. (5) Operability has been improved by simplification of the whole reactor system, by adoption of the passive safety systems, advanced automatic operation systems, and so on. (J.P.N.)

  2. Thermal-hydraulic modeling needs for passive reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  3. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  4. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  5. Conceptual design of the advanced marine reactor MRX

    Science.gov (United States)

    1991-02-01

    Design studies on the advanced marine reactors have been done continuously since 1983 at the Japan Atomic Energy Research Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type pressurized water reactor (PWR) built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study.

  6. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  7. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

  8. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  9. Preliminary design concept of an advanced integral reactor

    International Nuclear Information System (INIS)

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the reactor design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author)

  10. Removal heat extraction systems in advanced reactors

    International Nuclear Information System (INIS)

    The two main problems generally attributed to the electricity generation by nuclear power are the security of the facility and the radioactivity of the nuclear wastes, in a way that the only tasks of the European Commission on this matter are to make sure a high level of security in the facilities, as well as an adequate fuel and waste management. In this paper we discuss about the main lines in which the CIEMAT and the Polytechnic University of Valencia are working relative to the study of the passive working systems of the advanced designs reactors. (Author) 24 refs

  11. Conceptual design of the advanced marine reactor MRX

    International Nuclear Information System (INIS)

    Design studies on the advanced marine reactors have been done continuously since 1983 at JAERI in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type PWR, built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study. (author)

  12. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  13. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  14. ICONE-4: Proceedings. Volume 2: Advanced reactors

    International Nuclear Information System (INIS)

    The proceedings for this conference are contained in 5 volumes. This volume is divided into the following areas: advanced reactor requirements; advanced reactor design and analysis; arrangement and construction; specific reactor designs; demonstration testing; safety systems and analysis; component demonstration testing; advanced reactor containment design; licensing topics and updates; accelerator applications and spallation sources; and advanced reactor development. Separate abstracts were prepared for most papers in this volume

  15. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  16. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  17. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  18. Development of advanced nuclear reactors in Russia

    International Nuclear Information System (INIS)

    Several advanced reactor designs have been so far developed in Russia. The AES-91 and AES-92 plants with the VVER-1000 reactors have been developed at the beginning of 1990. However, the former design has been built in China and the latest which is certified meeting European Utility Requirements is being built in India. Moreover, the model VVER-1500 reactor with 50-60 MWd/t burn-up and an enhanced safety was being developed by Gidropress about 2005, excepting to be completed in 2007. But, this schedule has slipped in favor of development of the AES-2006 power plant incorporating a third-generation standardized VVER-1200 reactor of 1170 MWe. This is an evolutionary development of the well-proven VVER-1000 reactor in the AES-92 plant, with longer life, greater power and efficiency and its lead units are being built at Novovoronezh II, to start operation in 2012-13. Based on Atomenergoproekt declaration, the AES-2006 conforms to both Russian standards and European Utility Requirements. The most important features of the AES-2006 design are mentioned as: a design based on the passive safety systems, double containment, longer plant service life of 50 years with a capacity factor of 92%, longer irreplaceable components service life of 60 years, a 28.6% lower amount of concrete and metal, shorter construction time of 54 months, a Core Damage Frequency of 1x10-7/ year and lower liquid and solid wastes by 70% and 80% respectively. The presented paper includes a comparative analysis of technological and safety features, economic parameters and environmental impact of the AES-2006 design versus the other western advanced reactors. Since the Bushehr phase II NPP and several other NPPs are planning in Iran, such analysis would be of a great importance

  19. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  20. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  1. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  2. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  3. Next generation advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

  4. Reactor scram events in the updated PIUS 600 advanced reactor design

    International Nuclear Information System (INIS)

    The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission's preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams

  5. Passive cooling in modern nuclear reactors

    International Nuclear Information System (INIS)

    This paper presents some recent experimental results performed with the aim of understanding the mechanism of passive cooling. The AP 600 passive containment cooling system is simulated by an electrically heated vertical pipe, which is cooled by a naturally induced air flow and by a water film descending under gravity. The results demonstrate that although the presence of the water film improved the heat transfer significantly, the mode of heat transfer was very dependent on the experimental parameters. Preheating the water improved both film stability and overall cooling performance

  6. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  7. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  8. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  9. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  10. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  11. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  12. Performance of a passive emergency heat removal system of advanced reactors in two-phase flow and with high concentration of non-condensable

    International Nuclear Information System (INIS)

    The research and the development of passive emergency cooling systems are necessary for the new generation of thermo-nuclear systems. Some basic information on the operation of these systems require the research of some relative processes to the natural circulation, mainly in conditions of two-phase flow involving processes of condensation in the presence of non-condensable gases, because many found situations are new. The experimental facility called Bancada de Circulacao Natural (BCN) was used for the realization of tests with diverse concentrations of non-condensable and power levels. The non-condensable gas present in the circuit decreases the rate of heat transfer for the secondary of the heat exchanger, determining low efficiency of the heat exchanger. High concentration of non-condensable in the vapor condensation, determines negative pressure, and cause the inversion of the flow in the circuit. The initial concentration of non-condensable and the geometry of the circuit, in the inlet of the heat exchanger, determines the establishment of transitory with two-phase flow. The BCN was performed with the computational code of Analysis of Accidents and Thermal-Hydraulics RELAP5/MOD 3.3 and, the calculated values had been compared with the experimental data, presenting good agreement for small non-condensable concentrations. The values calculated for high concentrations of non-condensable had been satisfactory after the circuit to have reached the temperature of saturation in the electric heater. (author)

  13. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

    1999-03-01

    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  14. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  15. Passive heat-transfer means for nuclear reactors. [LMFBR

    Science.gov (United States)

    Burelbach, J.P.

    1982-06-10

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  16. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  17. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  18. Safety features of JAERI Passive Safety Reactor (JPSR)

    International Nuclear Information System (INIS)

    At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. The system consists of a Nuclear Steam Supply System (NSSS) and a steam and power conversion system. The NSSS has an inherent matching nature of core generation and heat removal rate and is a physically-slave system for the steam system. All of systems for the NSSS are built in the containment except for the air coolers of the final heat sink of the passive residual heat removal system. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system are adopted. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. Accordingly the reliability of the safety system and the normal operation system is improved, since most part of residual heat removal system is always working as auxiliary cooling system and a heat sink for normal operation system is 'safety class'. Analysis demonstrated the capability of passive cooling and depressurization

  19. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  20. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  1. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  2. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  3. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  4. Classification of systems for passive afterheat removal from reactor containment of nuclear power plant with water-cooled power reactor

    OpenAIRE

    Khaled, N.; D. V. Shevelev; A. S. Balashevsky

    2014-01-01

    A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  5. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  6. Performance Assessment of Passive Gaseous Provisions (PGAP). Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000 on the basis of IAEA General Conference resolution GC(44)/RES/21. INPRO helps to ensure the availability of sustainable nuclear energy in the 21st century and seeks to bring together all interested Member States - both technology holders and technology users - to consider joint actions to achieve desired innovations. To contribute to an international consensus on the definition of the reliability of passive systems that involve natural circulation, and on a methodology to assess this reliability, INPRO initiated a collaborative project on Performance Assessment of Passive Gaseous Provisions (PGAP) in 2007. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones, not only to enhance the operational safety of the reactors but also to mitigate the consequences of a severe accident should one occur. However, the reliability of passive safety systems is crucial and must be assessed before they are used extensively in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are a priori unknown. The functions of many passive systems are based on thermohydraulic principles, which until recently were considered as not being subject to any kind of failure. Hence, large and consistent efforts are required to quantify the reliability of such systems. Three participants from three INPRO Member States were involved in this collaborative project. Reliability methods for passive systems (RMPS) and assessment of passive system reliability (APSRA) methodologies were used by the participants to assess the performance and reliability of the passive decay heat removal system of the French gas cooled fast reactor design for station blackout and a loss of coolant accident combined with loss of off-site power, respectively. This publication presents the

  7. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  8. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  9. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  10. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  11. Passive Safety Small Reactor for Distributed Energy Supply

    Science.gov (United States)

    Ishida, Toshihisa; Sawada, Ken-Ichi; Odano, Naoteru

    The purpose of this paper is to study the core performance of passive safety small reactor for distributed energy supply by changing the heavy water (D2O) concentration in the mixed coolant together with the fuel pitch. The long core life with conditions of the excessive reactivity of 2 %Δk/k, the reactivity shutdown margin of 1 %Δk/k and the negative coolant temperature reactivity coefficient is attained for the case of D2O concentration of 60% with 10% enrichment gadolinia (Gd2O3) doped fuel rods. This D2O core has a shorter core life 4.14 years than the original light water (H2O) core 4.76 years, while it needs a larger core size. However, changing the D2O concentration on the way during the burn-up shows a possibility of extending more the core life than that of the original H2O core.

  12. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  13. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  14. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  15. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  16. Passive temperature monitoring in reactor environments using integrating thermal monitors

    International Nuclear Information System (INIS)

    A new passive temperature monitoring device, the Integrating Thermal Monitor (ITM) has been developed for measurements in reactor environments. The operation of the ITM is based on the thermal annealing behavior of latent particle tracks in Solid State Track Recorders (SSTRs). Several SSTR materials have been identified which have time/temperature annealing sensitivities for latent tracks which match the environments of the nuclear power plant measurement locations of interest. Of these, nine ITM materials have been calibrated at constant temperatures of 30, 40, 50, 60, 70, 80, 90, 100, 150, and 175 C for periods ranging from hours to more than 4 years. Detailed response measurements have been obtained for more than 600 ITM samples to establish a calibration data base against which ITM response can be measured. A model has been developed for ITM response as a function of time and temperature, and this model is used to interpret the response of ITMs deployed at power reactors. Methods for using ITMs to determine Arrhenius Equivalent Temperatures for equipment components operated in unknown temperature environments are described, and ITM performance to date in nuclear power plant applications is also summarized

  17. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  18. Series lecture on advanced fusion reactors

    International Nuclear Information System (INIS)

    The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

  19. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  20. New reactor programs from passive to pebble bed

    Energy Technology Data Exchange (ETDEWEB)

    Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    2002-07-01

    The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

  1. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    International Nuclear Information System (INIS)

    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  2. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Placco, Guilherme M.; Guimaraes, Lamartine N.F., E-mail: placco@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAV/DCTA) Sao Jose dos Campos, SP (Brazil); Santos, Rubens S. dos, E-mail: rsantos@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN -RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  3. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  4. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  5. Local AREA networks in advanced nuclear reactors

    International Nuclear Information System (INIS)

    The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

  6. The advanced liquid metal reactor actinide recycle system

    International Nuclear Information System (INIS)

    The current U.S. National Energy Strategy includes four key goals for nuclear policy: enhance safety and design standards, reduce economic risk, reduce regulatory risk, and establish an effective high-level nuclear waste program. The U.S. Department of Energy's Advanced Liquid Metal Reactor Actinide Recycle System is consistent with these objectives. The system has the ability to fulfill multiple missions with the same basic design concept. In addition to providing an option for long-term energy security, the system can be effectively utilized for recycling of actinides in light water reactor (LWR) spent fuel, provide waste management flexibility, including the reduction in the waste quantity and storage time and utilization of the available energy potential of LWR spent fuel. The actinide recycle system is comprised of (1) a compact liquid metal (sodium) cooled reactor system with optimized passive safety characteristics, and (2) pyrometallurgical metal fuel cycle presently under development of Argonne National Laboratory. The waste reduction of LWR spent fuel is accomplished by transmutation or fissioning of the longer-lived transuranic isotopes to shorter-lived fission products in the reactor. In this presentation the economical and environmental incentive of the actinide recycle system is addressed and the status of development including licensing aspects is described. 3 refs., 1 tab., 6 figs

  7. Trends vs. reactor size of passive reactivity shutdown and control performance

    International Nuclear Information System (INIS)

    In LMR [liquid metal reactor] concepts, favorable passive reactivity shutdown performance in response to ATWS [anticipated transient without scram] events has been shown to be achievable when measurable, integral reactivity parameters satisfy certain requirements. The trends in the integral reactivity parameters with reactor size for both oxide and metal fuel have been developed based on a data base of about two dozen reactor designs in the range 400 to 3,600 MWth. The general conclusion is that the favorable passive reactivity control features which accrue to the metallic-fueled reactors in the modular size range can be achieved as well in the larger commercial sizes

  8. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  9. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  10. Mirror Advanced Reactor Study interim design report

    International Nuclear Information System (INIS)

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design

  11. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  12. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  13. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  14. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  15. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  16. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  17. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  18. Final report-passive safety optimization in liquid sodium-cooled reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  19. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  20. ACR-1000TM - advanced Candu reactor design

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU ReactorTM- 1000 (ACR-1000TM) as an evolutionary advancement of the current CANDU 6TM reactor. This evolutionary advancement is based on AECL's in-depth knowledge of CANDU structures, systems, components and materials, gained during 50 years of continuous construction, engineering and commissioning, as well as on the experience and feedback received from operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. These innovations improve economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The Canadian nuclear reactor design evolution that has reached today's stage represented by the ACR-1000, has a long history dating back to the early 1950's. In this regard, Canada is in a unique situation, shared only by a very few other countries, where original nuclear power technology has been invented and further developed. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 and Phase 2 pre-project design reviews in December 2008 and August 2009, respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The final stage of the ACR-1000 design is currently underway and will be completed by fall of 2011, along with the final elements of the safety analyses and probabilistic safety analyses supporting the finalized design. The generic Preliminary Safety Analysis Report (PSAR) for the ACR-1000 was completed in September 2009. The PSAR demonstrates ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. (authors)

  1. Advanced Small Modular Reactor Economics Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

  2. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  3. Design study on the Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    Full text: The design study on the Advanced Recycling Reactor (ARR) has been conducted. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. International Nuclear Recycling Alliance (INRA) takes advantage of international experience and uses the design based on Japan Sodium-cooled Fast Reactor (JSFR) as reference for FOA studies of US DOE, because Japan has conducted R and Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. The targets of the ARR are to generate electricity while consuming fuel containing transuranics and to attain cost competitiveness with the similar sized LWRs. INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. INRA believes the scale-up factor of two is acceptable increase from manufacturing and licensing points of view. Major features of the ARR1 are the following: The reactor core is 70cm high and the volume fraction of fuel is approximately 32%. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45-51 kg/TWeh.Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop arrangement and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The ARR1 is co-located with a recycling facility. The overall plant facility arrangement is planned assuming to be constructed and installed in an inland area. The plant consists of a reactor building (including reactor auxiliary facilities and electrical/control systems), a turbine building, and a recycling building. The volume of the reactor building will be approximately 180,000 m3. The capital cost for the ARR1 and the ARR2 are

  4. Technical outline of a high temperature pool reactor with inherent passive safety features

    International Nuclear Information System (INIS)

    Many reactor designers world wide have successfully established technologies for very small reactors (less than 10 MWTH), and technologies for large power reactors (greater than 1000 MWTH), but have not developed small reactors (between 10 MWTH and 1000 MWth) which are safe, economic, and capable of meeting user technical, economic, and safety requirements. This is largely because the very small reactor technologies and the power reactor technologies are not amiable to safe and economic upsizing/downsizing. This paper postulates that new technologies, or novel combinations of existing technologies are necessary to the design of safe and economic small reactors. The paper then suggest a set of requirements that must be satisfied by a small reactor design, and defines a pool reactor that utilizes lead coolant and TRISO fuel which has the potential for meeting these requirements. This reactor, named LEADIR-PS, (an acronym for LEAD-cooled Integral Reactor, Passively Safe) incorporates the inherent safety features of the Modular High Temperature Gas Cooled Reactor (MWGR), while avoiding the cost of reactor and steam generator pressure vessels, and the safety concerns regarding pressure vessel rupture. This paper includes the description of a standard 200MW thermal reactor module based on this concept, called LEADIR-PS 200. (author)

  5. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  6. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    International Nuclear Information System (INIS)

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs

  7. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

  8. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  9. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  10. ASME Material Challenges for Advanced Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  11. Overview of the US program of controls for advanced reactors

    International Nuclear Information System (INIS)

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed

  12. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  13. Enhanced passive safety features against ATWS of fast breeder reactors with capabilities of MA incineration

    Energy Technology Data Exchange (ETDEWEB)

    Ninokata, Hisashi; Sawada, Tetsuo; Sato, Manabu [Tokyo Institute of Technology (Japan)] [and others

    1997-12-01

    The paper gives an outline of the general and simple reactivity correlation method to identify the region of the major design parameters that assures power stabilization and passive shutdown of sodium-cooled large fast reactors under ATWS conditions. Based on the model developed, general design guidelines are shown that enhance passive capabilities being aimed at preventing sodium boiling and fuel failures in the events of ULOF and UTOP. Discussions extend to the influences of minor actinides loading in the core onto the passive safety features. 6 refs., 1 fig., 1 tab.

  14. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-31

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  15. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    International Nuclear Information System (INIS)

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  16. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    OpenAIRE

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years. However, the design of nuclear reactors can be improved significantly in terms of safety, by designing reactors with fully passive safety systems, and sustainability, by making more efficient use of natu...

  17. Trends in advanced reactor development and the role of the IAEA

    International Nuclear Information System (INIS)

    This report discusses advanced reactors are being developed for all principal reactor types, i.e. the light and heavy water-cooled reactors, the liquid-metal-cooled reactors and the gas-cooled reactors. Some of these developments are primarily of an evolutionary nature, i.e. they represent improvements in component and system technology, and in construction and operating practices as a result of experience gained with presently operating plants. Other developments are also evolutionary but with some incorporation of innovative features such as providing passive systems for assuring continuous cooling for removal of decay heat from the reactor core. If there is a revival of nuclear power, which may be dictated by ecological and economical factors, advanced reactors now being developed could help to meet the large demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, has promoted international information exchange and international cooperation between all countries with their own advanced nuclear power programmes and has offered assistance to countries with an interest in exploratory or research programmes. In the future the IAEA could play an even more-important role

  18. Development Program of the Advanced HANARO Reactor in Korea

    International Nuclear Information System (INIS)

    The development program of an advanced HANARO (AHR) reactor started in Korea to keep abreast of the increasing future demand, from both home and abroad, for research activities. This paper provides a review of the status of research reactors in Korea, the operating experience of the HANARO, the design principles and preliminary features of an advanced HANARO reactor, and the specific strategy of an advanced HANARO reactor development program. The design principles were established in order to design a new multi-purpose research reactor that is safe, economically competitive and technically feasible. These include the adaptation of the HANARO design concept, its operating experience, a high ratio of flux to power, a high degree of safety, improved economic efficiency, improved operability and maintainability, increased space and expandability, and ALARA design optimization. The strategy of an advanced HANARO reactor development program considers items such as providing a digital advanced HANARO reactor in cyber space, a method for the improving the design quality and economy of research reactors by using Computer Integrated Engineering, and more effective advertising using diverse virtual reality. This development program will be useful for promoting the understanding of and interest in the operating HANARO as well as an advanced HANARO reactor under development in Korea. It will provide very useful information to a country that may need a research reactor in the near future for the promotion of public health, bio-technology, drug design, pharmacology, material processing, and the development of new materials. (author)

  19. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  20. Advanced nuclear reactor public opinion project

    International Nuclear Information System (INIS)

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions

  1. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  2. Advanced Small Modular Reactor Economics Model Development

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

  3. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  4. A Joint Report on PSA for New and Advanced Reactors

    International Nuclear Information System (INIS)

    This report addresses the application of Probabilistic Safety Assessment (PSA) to new and advanced nuclear reactors. As far as advanced reactors are concerned, the objectives were to characterize the ability of current PSA technology to address key questions regarding the development, acceptance and licensing of advanced reactor designs, to characterize the potential value of advanced PSA methods and tools for application to advanced reactors, and to develop recommendations for any needed developments regarding PSA for these reactors. As far as the design and commissioning of new nuclear power plants is concerned, the objectives were to identify and characterize current practices regarding the role of PSA, to identify key technical issues regarding PSA, lessons learned and issues requiring further work; to develop recommendations regarding the use of PSA, and to identify future international cooperative work on the identified issues. In order to reach these objectives, questionnaires had been sent to participating countries and organisations

  5. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    International Nuclear Information System (INIS)

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  6. A modern control room for Indian Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a next generation nuclear power plant being developed by Bhabha Atomic Research Centre, India. AHWR is a vertical, pressure tube type, heavy-water-moderated, boiling light-water-cooled, innovative reactor, relying on natural circulation for core cooling in all operating and accident conditions. In addition, it incorporates various passive systems for decay heat removal, containment cooling and isolation. In addition to the many passive safety features, AHWR has state of the art I and C architecture based on extensive use of computers and networking. In tune with the many advanced features of the reactor, a centralized modern control room has been conceived for operation and monitoring of the plant. The I and C architecture enables the implementation of a fully computerised operator friendly control room with soft Human Machine Interfaces (HMI). While doing so, safety has been given due consideration. The control and monitoring of AHWR systems are carried out from the fully computer-based operator interfaces, except safety systems, for which only monitoring is provided from soft HMI. The control of the safety systems is performed from dedicated hardwired safety system panels. Soft HMI reduces the number of individual control devices and improves their reliability. The paper briefly describes the I and C architecture adopted for the AHWR plant along with the interfaces to the main and backup control rooms. There are many issues involved while introducing soft HMI based operator interfaces for Nuclear Power Plants (NPP) compared to the conventional plants. Besides discussing the implementation issues, the paper elaborates the design considerations that have undergone in the design of various components in the main control room especially operator workstations, shift supervisor console, safety system panels and large display panels. Mainly task based displays have been adopted for the routine operator interactions of the plant

  7. Municipal waste stabilization in a reactor with an integrated active and passive aeration system.

    Science.gov (United States)

    Kasinski, Slawomir; Slota, Monika; Markowski, Michal; Kaminska, Anna

    2016-04-01

    To test whether an integrated passive and active aeration system could be an effective solution for aerobic decomposition of municipal waste in technical conditions, a full-scale composting reactor was designed. The waste was actively aerated for 5d, passively aerated for 35 d, and then actively aerated for 5d, and the entire composting process was monitored. During the 45-day observation period, changes in the fractional, morphological and physico-chemical characteristics of the waste at the top of the reactor differed from those in the center of the reactor. The fractional and morphological analysis made during the entire process of stabilization, showed the total reduction of organic matter measured of 82 wt% and 86 wt% at the respective depths. The reduction of organic matter calculated using the results of Lost of Ignition (LOI) and Total Organic Carbon (TOC) showed, respectively, 40.51-46.62% organic matter loss at the top and 45.33-53.39% in the center of the reactor. At the end of the process, moisture content, LOI and TOC at the top were 3.29%, 6.10% and 4.13% higher, respectively, than in the center. The results showed that application of passive aeration in larger scale simultaneously allows the thermophilic levels to be maintained during municipal solid waste composting process while not inhibiting microbial activity in the reactor.

  8. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  9. Passive and inherent safety technologies for light-water nuclear reactors

    International Nuclear Information System (INIS)

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

  10. Natural heat transfer augmentation in passive advanced BWR plants

    International Nuclear Information System (INIS)

    In the European Simplified Boiling Water Reactor (ESBWR), the long-term post-accident containment pressure is determined by the combination of non condensable gas pressure and steam pressure in the wet well gas space. Since there are no active systems for heat removal in the wet well, energy transmitted to the wet well gas space, by a variety of means, must be removed by passive heat transfer to the walls and suppression pool (SP). The cold suppression pool located below the hotter gas space provides a stable configuration in which convection currents are suppressed thus limiting heat and mass transfer between the gas space and pool. However, heat transfer to the walls results in natural circulation currents that can augment the heat and mass transfer to the pool surface. Using a simplified model, parametric studies are carried out to show that augmentation of the order of magnitude expected can significantly impact the heat and mass transfer to the pool. Additionally a review of available literature in the area of augmentation and mixed convection of this type is presented and indicates the need for additional experimental work in order to develop adequate models for heat and mass transfer augmentation in the configuration of a BWR suppression pool. (author)

  11. Pressure suppression pool mixing in passive advanced BWR plants

    International Nuclear Information System (INIS)

    In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data

  12. Tests on a prototype of the Passive Fission Gas Monitor for failed detection (PRISM reactor)

    International Nuclear Information System (INIS)

    The Passive Diffusion Fission Gas Monitor PDFGM is mounted on the PRISM reactor head and extends into the cover gas Region where it determines the presence of radioactive fission gases (Kr, Xe, and so on) released from failed fuel pins. It contains a steel diffusion column that is closed at the upper end but opened to the cover gas at its lower end. The upper portion of the column is located in the field of view of a collimated gamma detector which is shielded from the remainder of the cover gas and of the sodium pool below. Heaters are provided to obtain a uniform axial temperature in the gas column and to minimize the potential for natural convection currents. In this way, the molecular diffusion can be established based on the fission gas concentration gradients along the column length. This is an advanced solution in comparison with current devices based on active components (pumps, filters, and so on). The experimental results on a prototype of PDFGM and their interpretation will be presented in this paper. (author)

  13. Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

    International Nuclear Information System (INIS)

    In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

  14. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  15. Passive Safety Optimization in Liquid Sodium-Cooled Reactors

    International Nuclear Information System (INIS)

    The governing equations and the solutions schemes are developed for the three-dimensional thermal-hydraulic model.A detailed constitutive relations are also developed through the analysis with a CFD code. The developed model is able to obtain a detailed thermal hydraulic information in a subassemlby of a liquid metal-cooled reactor core. The model has been integrated with the system analysis codes SASSYS-1 and SSC-K to be validated for the SHRT-17 test performed in the EBR-II reactor. The baseline analyses were performed with the customary thermal-hydraulic model and with the new model for the reference design of KALIMER-150. The results point out that increased detail in the thermal-hydraulic analysis must be matched by increased detail in the reactivity feedback modeling, especially the radial core expansion model, in order to provide distinguishable differences. The analyses performed to identify certain key safety features have indicated that the adjustment of core restraint system is able to provide beneficial bending of hexcans, resulting in enhanced negative reactivity feedback in unprotected accidents. Tasks were performed to identify and assess the implications for plant safety of proposed specific approaches for reducing capital and operating costs of next generation sodium-cooled fast reactors. A new plant design of the sodium-cooled reactor concept is developed utilizing a gas turbine Brayton cycle, which uses supercritical carbon dioxide (S-CO2) as the working fluid. In addition, the design innovations are incorporated in modular sodium-to-supercritical carbon dioxide heat exchangers that enable the traditional intermediate heat transport circuit to be eliminated. Several evaluations and analyses for the safety design and efficiency suggest that the developed system is safe and cost-effective. Test plans are developed for the measurement of phenomenological data describing freezing of molten metallic fuel, melt relocation and interaction with steel

  16. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  17. A severe accident analysis for the system-integrated modular advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Gunhyo; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2015-03-15

    The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

  18. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  19. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  20. ORNL R and D on advanced small and medium power reactors: Selected topics

    Energy Technology Data Exchange (ETDEWEB)

    White, J.D.; Trauger, D.B.

    1988-01-01

    From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, and assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable RandD would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current RandD efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described. 13 refs., 1 fig.

  1. ORNL R and D on advanced small and medium power reactors: Selected topics

    International Nuclear Information System (INIS)

    From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, and assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable RandD would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current RandD efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described. 13 refs., 1 fig

  2. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  3. Utility Leadership in Defining Requirements for Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    It is appropriate, based on twenty five years of operating experience, that utilities take a position of leadership in developing the technical design and performance requirements for the next generations of nuclear electric generating plants. The U. S. utilities, through the Electric Power Research Institute, began an initiative in 1985 to develop such Utility requirements. Many international Utility organizations, including Korea Electric Power Corporation, have joined as full participants in this important Utility industry initiative. In light of the closer linkage among countries of the world due to rapid travel and telecommunications, it is also appropriate that there be international dialogue and agreement on the principal standards for nuclear power plant acceptability and performance. The Utility/EPRI Advanced Light Water Reactor Program guided by the ALRR Utility Steering Committee has been very successful in developing these Utility requirements. This paper will summarize the state of development of the ALRR Utility Requirements for Evolutionary Plants, recent developments in their review by the U. S. Nuclear Regulatory Commission, resolution of open issues, and the extension of this effort to develop a companion set of ALRR Utility Requirements for plants employing passive safety features

  4. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  5. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO2) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

  6. Shielding considerations for advanced space nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  7. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  8. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  9. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  10. China Advanced Research Reactor Project Progress in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    2011, China Advanced Research Reactor (CARR) Project finished the B stage commissioning and resolved the relative technical problems. Meanwhile, the acceptance items and the cold neutron source were carrying out.

  11. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  12. Westinghouse Small Modular Reactor passive safety system response to postulated events

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M. C.; Wright, R. F. [Westinghouse Electric Company, 600 Cranberry Woods Drive (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

  13. Westinghouse Small Modular Reactor passive safety system response to postulated events

    International Nuclear Information System (INIS)

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000R reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce

  14. Analysis of passive residual heat removal system of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The passive residual heat removal system plays an important role for the inherent safety of high temperature gas-cooled reactor (HTGR). The thermal hydraulic calculation method for the residual heat removal system of HTGR was introduced. The operating temperatures of the residual heat removal system at different residual heat powers and different environmental temperatures were calculated. The containment concrete temperature was numerically simulated. The results show that the highest concrete temperature is acceptable. (authors)

  15. Passive Safety of the STAR-LM HLMC Natural Convection Reactor

    International Nuclear Information System (INIS)

    The STAR-LM 300 to 400 MWt class modular, factory fabricated, fully transportable, proliferation resistant, autonomous, reactor system achieves passive safety by taking advantage of the intrinsic benefits of inert lead-bismuth eutectic heavy liquid metal coolant, 100+% natural circulation heat transport, a fast neutron spectrum core utilizing high thermal conductivity transuranic nitride fuel, redundant passive air cooling of the outside of the guard/containment vessel driven by natural circulation, and seismic isolation where required by site conditions. Postulated loss-of-heat sink without scram, overcooling without scram, and unprotected transient overpower accidents are analyzed for the 300 MWt STAR-LM design using a coupled thermal hydraulics-neutron kinetics plant dynamics analysis computer code. In all cases, STAR-LM is calculated to exhibit passive safety with peak cladding and coolant temperatures remaining within the existing database for lead-bismuth eutectic coolant and ferritic steel core materials. (authors)

  16. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    Science.gov (United States)

    Zweibaum, Nicolas

    The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). (Abstract shortened by UMI.).

  17. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    Science.gov (United States)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  18. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  19. Advanced Passive Microwave Radiometer Technology for GPM Mission

    Science.gov (United States)

    Smith, Eric A.; Im, Eastwood; Kummerow, Christian; Principe, Caleb; Ruf, Christoper; Wilheit, Thomas; Starr, David (Technical Monitor)

    2002-01-01

    An interferometer-type passive microwave radiometer based on MMIC receiver technology and a thinned array antenna design is being developed under the Instrument Incubator Program (TIP) on a project entitled the Lightweight Rainfall Radiometer (LRR). The prototype single channel aircraft instrument will be ready for first testing in 2nd quarter 2003, for deployment on the NASA DC-8 aircraft and in a ground configuration manner; this version measures at 10.7 GHz in a crosstrack imaging mode. The design for a two (2) frequency preliminary space flight model at 19 and 35 GHz (also in crosstrack imaging mode) has also been completed, in which the design features would enable it to fly in a bore-sighted configuration with a new dual-frequency space radar (DPR) under development at the Communications Research Laboratory (CRL) in Tokyo, Japan. The DPR will be flown as one of two primary instruments on the Global Precipitation Measurement (GPM) mission's core satellite in the 2007 time frame. The dual frequency space flight design of the ERR matches the APR frequencies and will be proposed as an ancillary instrument on the GPM core satellite to advance space-based precipitation measurement by enabling better microphysical characterization and coincident volume data gathering for exercising combined algorithm techniques which make use of both radar backscatter and radiometer attenuation information to constrain rainrate solutions within a physical algorithm context. This talk will discuss the design features, performance capabilities, applications plans, and conical/polarametric imaging possibilities for the LRR, as well as a brief summary of the project status and schedule.

  20. Advanced Passive Reactors : Leading The U. S. Nuclear Renaissance

    International Nuclear Information System (INIS)

    Twenty-one years have passed since Korea Electric Power Corporation and Westinghouse announced plans to build Kori 1. Today, Korea's nuclear program is one of the most successful in the world. The electricity generated from Kori 1 and eight other nuclear plants has helped to spark the remarkable growth and transformation of Korea into a modern industrial power. Westinghouse is proud to have been Korea's partner on six of those plants. It the past is the bast prophet of the future, then you and your countrymen should certainly be excited by your future. Korean industry is poised to continue its steady growth, and that means continued growth for your nuclear industry. Currently, the U. S. nuclear industry is experiencing a similar mood of excitement. In fact, it would be necessary to go almost all the way back to the beginning of the birth of the Korean nuclear industry, in 1969, to find a time when the future of nuclear power in the United States looked as bright as it does today. Part of our excitement stems from the welcome prospect of growth. In recent years, there has not been a market for new nuclear plants in the United States. Utilities either had excess capacity or were building plants they had ordered before 1974. For example, between 1980 and 1989, U. S. utilities completed 46 large nuclear units, but didn't order a single new one in that time. Since 1983, however, strong economic growth in the United States has caused the demand for electric power to grow about twice as fast as utilities had projected. Today, utilities will need to order new busload plants. When they do, utilities won't want technology developed 20 years ago. They'll be looking for plants that can meet the environmental, economic, and safety standards of the 21st century

  1. Design and development of high temperature heat pipes and thermosiphons for passive heat removal from compact high temperature reactor

    International Nuclear Information System (INIS)

    Compact High Temperature Reactor (CHTR) is 100 kWth, lead-bismuth eutectic (LBE) cooled reactor having several advanced passive safety features to enable its operation as compact power pack. It will also facilitate demonstration of technologies for high temperature process heat applications. In CHTR heat is transferred from primary to secondary side by means of high temperature heat pipes. Heat pipes are also employed to remove heat under postulated accident scenarios. Thus, reliable operation of heat pipes is essential for the safe working of the reactor. In this respect, computer codes have been developed for design and simulation of high temperature heat pipes. This includes design codes using empirical correlations as well as simplified FEM models for system level analysis. To verify the operation of these heat pipes under various steady state and transient conditions full CFD analysis is essential. This has been done by using a commercial CFD code by incorporating user defined functions (UDFs) which address the saturated nature of the vapour phase and the vapour wick interface conditions. A three dimensional transient numerical model has been developed to predict the vapor core, wall temperatures, vapor pressure, and vapor velocity in the screen mesh wick of sodium heat pipe. This thesis will give an outline of all the developed models and compared the predicted results against the experimental data. (author)

  2. Development of inherent core technologies for advanced reactor

    International Nuclear Information System (INIS)

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  3. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  4. Experimental studies on in-bundle ECCS injection for Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) being designed at BARC is an innovative reactor with Thorium utilization as its major objective. It has many advanced passive safety features. One such feature is passive injection of emergency coolant after postulated Loss of Coolant Accident (LOCA). A novel feature of this injection scheme is that the injection does not take place in the header/plenum as in other reactors, but directly in to the bundle. For this purpose, the fuel cluster incorporates a central water rod which communicates with the ECCS header. The water rod extends along full length of the fuel cluster. In event of LOCA in the Main Heat Transport (MHT) system, ECC water flows from the accumulator to the water rod through ECCS header. The water flows into the bundle through holes in the water rod. The AHWR fuel cluster has fuel pins arranged in three concentric rings (of 12, 18 and 24 pins) around the central rod. While it is ensured that water does reach the fuel cluster, whether it reaches the outer ring of pins is needs investigation as the pins are closely spaced (1-3 mm gap between adjacent rods). The objective of the present experiments is to determine under what conditions (ECC flow and decay heat), the ECC water is able to rewet and cool all the fuel pins. The experiments have been done in a short, instrumented fuel bundle simulating the geometry of the AHWR fuel cluster

  5. Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

    International Nuclear Information System (INIS)

    Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

  6. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  7. Advanced Reactors Thermal Energy Transport for Process Industries

    Energy Technology Data Exchange (ETDEWEB)

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  8. Decay heat removal in pool type fast reactor using passive systems

    Energy Technology Data Exchange (ETDEWEB)

    Parthasarathy, U. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Sundararajan, T. [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Balaji, C., E-mail: balaji@iitm.ac.in [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Velusamy, K.; Chellapandi, P.; Chetal, S.C. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. Black-Right-Pointing-Pointer Calculations confirm adequacy of natural convection in decay heat removal. Black-Right-Pointing-Pointer Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the

  9. Decay heat removal in pool type fast reactor using passive systems

    International Nuclear Information System (INIS)

    Highlights: ► Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. ► Calculations confirm adequacy of natural convection in decay heat removal. ► Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement

  10. US advanced liquid metal reactor design and safety enhancements through teamwork with the NRC

    International Nuclear Information System (INIS)

    This paper presents a cost effective approach to providing electricity needs of the United States in the early 21st century through an integrated Advanced Liquid Metal Reactor System (ALMRS). The concept incorporates innovative design and passive safety features to achieve favourable safety and economics simultaneously. The synergistic components of the ALMRS, under development by the United States (US) Department of Energy (DOE), are described. A key aspect of the programme is continuing interaction with US Nuclear Regulatory Commission (NRC) toward standard design certification. This interaction, and the results to date, are also addressed. (author)

  11. Utility leadership in reopening the nuclear option with advanced light water reactors

    International Nuclear Information System (INIS)

    Since 1981, the Electric Power Research Institute (EPRI) has been pursing the development of the advanced light water reactor (ALWR). The ALWR Program is comprised of five phases and are described in the paper. In order to meet the anticipated baseline power generation requirements in the US, the Nuclear Power Oversight Committee (NPOC) has developed a strategic plan for ALWR implementation in order to regain the nuclear option in the United States. The paper also covers the policies behind the utility requirements, the status of ALWR developments in the United States, the electricity demands during the period 1990-2010, and some of the innovative features of the passive plants presently under design

  12. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  13. A dynamic model of a passively cooled small modular reactor for controller design purposes

    Energy Technology Data Exchange (ETDEWEB)

    Arda, Samet E., E-mail: s.e.arda@asu.edu; Holbert, Keith E., E-mail: holbert@asu.edu

    2015-08-15

    Highlights: • A mathematical dynamic model is developed for a passively cooled small modular reactor. • Reactor response associated single-phase natural circulation is analyzed. • A moving boundary model for a helical-coil steam generator is analyzed. • Dynamic responses of the overall model to representative perturbations are evaluated. • This compact model can be utilized for control system design. - Abstract: An analytical dynamic model for a passively cooled small modular reactor (SMR) is developed using a state-variable lumped parameter approach. Reactor power is represented by the generation time formulation of the point kinetics equations with a single combined neutron precursor group. The heat transfer process in the core is described via an overall heat transfer coefficient by defining two coolant lumps paired to a single fuel lump. In addition, a thermal–hydraulics model for single-phase natural circulation is incorporated. For the helical-coil steam generator, a moving-boundary model including subcooled, two-phase, and superheated regions is utilized. Finally, the hot leg riser and downcomer regions are expressed by first-order lags. The performance of the overall system described by ordinary differential equations (ODEs) is evaluated by the Simulink dynamic environment and directly using a MATLAB ODE solver recommended for stiff systems. Simulation results based on NuScale SMR design data show that the initial steady-state values for 100% power are within range of the design data and the model can predict the system dynamics due to typical perturbations, e.g., control rod movement and change in feedwater mass flow rate and temperature. The model developed in this work can be utilized as a foundation for designing and testing a suitable control algorithm for reactor thermal power.

  14. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  15. Fusion reactor design towards radwaste minimum with advanced shield material

    International Nuclear Information System (INIS)

    A new concept of fusion reactor design is proposed to minimize the radioactive waste of the reactor. The main point of the concept is to clear massive structural components located outside the neutron shield from regulatory control. The concept requires some reinforcement of shielding with an advanced shield material such as a metal hydride, detriation, and tailoring of a detrimental element from the superconductor. Our assessment confirmed a large impact of the concept on radwaste reduction, in that it reduces the radwaste fraction of a fusion reactor A-SSTR2 from 92 wt.% to 17 wt.%. (author)

  16. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  17. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  18. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

    2009-10-15

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  19. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    Science.gov (United States)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  20. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  1. Neutronic challenges of advanced boiling water reactor designs

    International Nuclear Information System (INIS)

    The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)

  2. Conceptual design of a passive, inherently safe emergency shutdown rod for high-temperature reactor applications

    International Nuclear Information System (INIS)

    The concept of a passive, inherently safe, and fail-safe design for an emergency control rod is presented. The functioning of the rod is based solely on inexorable physical laws. The operation of the rod in its emergency function does not require the intervention of a human operator, nor does it rely on any signal from a monitoring or safety system. Although the concept could be applicable to a variety of reactors (provided a normal temperature range is specified), in this paper, the concept is applied to the emergency shutdown of a pebble-bed reactor. The preliminary study presented here demonstrates that the proposed Electro-Magnetic Optimally Scramming Control Rod (EM-OSCR) naturally operates when needed. The rod is held out of the core region by the force of an electromagnet. The force is generated by a current carried by a conductor, a portion of which passes near or through the reactor core region. When the temperature in the conductor increases because of an increase in temperature in the reactor, the conductor resistivity increases. This, in turn, leads to a current decrease. When the current decreases below the level necessary to hold the rod up, the rod is released and it falls into the core under the effect of gravity. (author)

  3. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  4. The study of steam explosions in nuclear systems. Advanced Reactor Severe Accident Program

    International Nuclear Information System (INIS)

    This report presents an overview of the steam explosion issue in nuclear reactor safety and our approach to assessing it. Key physics, models, and computational tools are described, and illustrative results are presented for ex-vessel steam explosions in an open pool geometry. An extensive set of appendices facilitate access to previously reported work that is an integral part of this effort. These appendices include key developments in our approach, key advances in our understanding from physical and numerical experiments, and details of the most advanced computational results presented in this report. Of major significance are the following features: A consistent two-dimensional treatment for both premixing and propagation which in practical settings are ostensibly at least two-dimensional phenomena; experimental demonstration of voiding and microinteractions which represent key behaviors in premixing and propagation respectively; demonstration of the explosion venting phenomena in open pool geometries which, therefore, can be counted on as a very important mitigative feature; and introduction of the idea of penetration cutoff as a key mechanism prohibiting large-scale premixing in usual ex-vessel situations involving high pour velocities and subcooled pools. This report is intended as an overview and is to be followed by code manuals for PM-ALPHA and ESPROSE.m, respective verification reports, and application documents for reactor-specific applications. The applications will employ the Risk Oriented Accident Analysis Methodology (ROAAM) to address the safety importance of potential steam explosions phenomena in evaluated severe accidents for passive Advanced Light Water Reactors (ALWRs)

  5. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  6. A wall-crawling robot for reactor vessel inspection in advanced reactors

    International Nuclear Information System (INIS)

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

  7. A wall-crawling robot for reactor vessel inspection in advanced reactors

    International Nuclear Information System (INIS)

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

  8. Advances in passive neutron instruments for safeguards use

    International Nuclear Information System (INIS)

    Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the 240Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ''known alpha'' and ''known multiplication'' methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems

  9. Advances in passive neutron instruments for safeguards use

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Krick, M.S.; Langner, D.G.; Miller, M.C.; Stewart, J.E.

    1994-02-01

    Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the {sup 240}Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ``known alpha`` and ``known multiplication`` methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems.

  10. Astronaut Kevin Chilton works with advanced cell reactor

    Science.gov (United States)

    1994-01-01

    Astronaut Kevin P. Chilton, pilot, works with an advanced cell reactor, which incorporated the first ever videomicroscope, on the Space Tissue Loss (STL-B) experiment on the Space Shuttle Endeavour's middeck. This experiment studied cell growth during the STS-59 mission.

  11. China Advanced Research Reactor Project Progress in 2012

    Institute of Scientific and Technical Information of China (English)

    ZHAO; Tie-jun

    2012-01-01

    <正>In 2012, all the commissioning for the China Advanced Research Reactor (CARR) had been finished and the diffraction pattern had been successfully obtained on the neutron scattering spectrometer. Meanwhile, the cold neutron source project and the acceptance items of CARR project had been carrying out.

  12. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  13. Identification of improvements of advanced light water reactor concepts

    International Nuclear Information System (INIS)

    The scope of this report is to identify the improvement of reactor developments with respect to reactor safety. This includes the collection of non-proprietary information on the description of the advanced design characteristics, especially summary design descriptions and general publications. This documentation is not intended to include a safety evaluation of the advanced concepts; however, it is structured in such a way that it can serve as a basis for a future safety evaluation. This is taken into account in the structure of the information regarding the distinction of the various concepts with respect to their 'advancement' and the classification of design characteristics according to some basic safety aspects. The overall description concentrates on those features which are relevant to safety. Other aspects, such as economy, operational features, maintenance, the construction period, etc...are not considered explicitly in this report

  14. Metal fire implications for advanced reactors. Part 1, literature review.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-10-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

  15. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith

    2013-09-01

    During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

  16. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  17. Proceedings of an international workshop on passive system reliability - A challenge to reliability engineering and licensing of advanced nuclear power plants

    International Nuclear Information System (INIS)

    premature at this time until a firm order is made and a request for licensing is received for one of the new reactors under design. It is apparent from the discussions at this workshop and others recently that future reactors will require a more advanced interpretation of defence-in-depth principles integrated with risk-informed thinking (i.e.; risk insights). In relation to international activities ongoing and planned on passive systems, two research programmes were discussed as follows: - Reliability Methods for Passive Systems (RMPS) - This is project underway by the European Commission with the objective to propose a specific methodology to assess the reliability of thermal hydraulic passive systems. - Proposed IAEA Co-ordinated Research Programme (CRP) on Natural Circulation Phenomena, modelling and reliability of passive systems which utilise natural circulation. The main purpose was to provide a forum for the exchange of information on the technical issues associated with assessing the reliability of passive systems in the context of nuclear safety, regulatory practices and probabilistic safety analysis. The workshop will discuss the state of the art of the topic, and will treat all types of passive systems, i.e. structural components, systems based on working fluids etc. A special emphasis will be put on the reliability of the systems based on thermal hydraulics, for which the methods are still in developing phase. Issues and discussions topics will include the lessons learned from designing passive systems, developing methodologies, performed studies, field experience with passive systems and need for future development. Participants will have the opportunity to meet their counterparts from other countries and organisations to discuss technical issues and problems associated with modelling and quantifying passive system reliability

  18. Johnson Noise Thermometry for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  19. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  20. Mirror Advanced Reactor Study (MARS) final report summary

    International Nuclear Information System (INIS)

    The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory

  1. Economics of advanced light water reactors - Recent update

    International Nuclear Information System (INIS)

    This paper includes recently updated analyses of the economic prospects of advanced light water reactors (ALWRs) during the decade of the 1990s. United Engineers and Constructors (UE and C) has performed engineering economic analyses related to ALWRs over the last 5 yr using both target economics and detailed cost-estimating methodologies. It has been found through such cost comparisons that properly designed and constructed ALWRs should cost less than the target cost figures listed above and significantly less than the pressurized water reactor better experience reference LWR plant cost

  2. Technology advances in active and passive microwave sensing through 1985

    Science.gov (United States)

    Barath, F. T.

    1977-01-01

    As a result of a growing awareness by the remote sensing community of the unique capabilities of passive and active microwave sensors, these instruments are expected to grow in the next decade in numbers, versatility and complexity. The Nimbus-G and Seasat-A Scanning Multichannel Microwave Spectrometer (SMMR), the Seasat-A radar altimeter, scatterometer and synthetic aperture radar represent the first systematic attempt at exploring a wide variety of applications utilizing microwave sensing techniques and are indicators of the directions in which the pertinent technology is likely to evolve. The trend is toward high resolution multi-frequency imagers spanning wide frequency ranges and wide swaths requiring sophisticated receivers, real-time data processors and most importantly, complex antennas.

  3. Advanced reactor instrumentation and control reliability and risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, R.; Gunther, W.; Valente, J.; Azarm, M.A.

    1991-12-31

    Advanced nuclear power reactors will used different approaches to achieving a higher level of safety than the first generation. One approach used the technological developments in computation and electronics in the form of digital instrumentation and control (I&C) to enhance the reliability, and accuracy of information for plant control, responding to the information, and controlling the plant and its systems under normal and upset environments in various states of degradation. Evaluating the reliability and safety of advanced I&C systems requires determining the reliability of the I&C used in the advanced reactors which involves distributed processing, data pile-up, interactive systems, the man-machine interface, various forms of automatic control, and systems interactions. From these analyses will come an understanding of the potential of the new I&C, and protection from its vulnerabilities to enhance the safe operation of the new plants. Technological, safety, reliability, and regulatory issues associated with advanced I&C for the new reactors are discussed herein. The issues are presented followed by suggested approaches to their resolution.

  4. Advanced reactor instrumentation and control reliability and risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, R.; Gunther, W.; Valente, J.; Azarm, M.A.

    1991-01-01

    Advanced nuclear power reactors will used different approaches to achieving a higher level of safety than the first generation. One approach used the technological developments in computation and electronics in the form of digital instrumentation and control (I C) to enhance the reliability, and accuracy of information for plant control, responding to the information, and controlling the plant and its systems under normal and upset environments in various states of degradation. Evaluating the reliability and safety of advanced I C systems requires determining the reliability of the I C used in the advanced reactors which involves distributed processing, data pile-up, interactive systems, the man-machine interface, various forms of automatic control, and systems interactions. From these analyses will come an understanding of the potential of the new I C, and protection from its vulnerabilities to enhance the safe operation of the new plants. Technological, safety, reliability, and regulatory issues associated with advanced I C for the new reactors are discussed herein. The issues are presented followed by suggested approaches to their resolution.

  5. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  6. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  7. Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-04-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

  8. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  9. Direct-contact condensation regime map for core makeup tank of passive reactors

    International Nuclear Information System (INIS)

    The condensation regime map in the core makeup tank of passive reactors is experimentally investigated. The condensation regimes identified through the experiments are divided into three distinct ones: sonic jet, subsonic jet, and steam cavity. The steam cavity regime is a unique regime of downward injection with the present geometry not previously observed in other experiments. The condensation regime map is constructed using Froude number and Jacob number. It turns out that the buoyancy force has a large influence on the regime transition because the regime map using the Froude number better fits data with different geometries than other dimensionless parameters. Simple correlations for the regime boundaries are proposed using the Froude number and the Jacob number

  10. Johnson Noise Thermometry for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Britton Jr, Charles L [ORNL; Roberts, Michael [ORNL; Bull, Nora D [ORNL; Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2012-10-01

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  11. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    International Nuclear Information System (INIS)

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300 C and 900 C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  12. Study of Pu consumption in Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology

  13. Modelling activities of experimental facilities related to advanced reactors. Considerations on 1D/3D issues

    International Nuclear Information System (INIS)

    The state of art of modelling activities related to integral experimental facilities of advanced passive reactors show to date important open items. The main advantage of using 1D plant codes is the capability of simulating the full interaction between components traditionally correctly modelled (condensers, heat exchangers, pipes and vessels) and other components for which codes are not 100% suitable (pools and containments). Polytechnical University of Catalonia (UPC) and Polytechnical University of Valencia (UPV) cooperated with other European research organizations in the 'Technology Enhancement for Passive Safety Systems' (TEPSS) project, within the European Fourth Framework Programme. It was a task of both Universities to supply analytical support of PANDA tests. The paper deals with the 1D/3D discussion in the framework of modelling activities related to integral passive facilities like PANDA. It starts choosing reference tests among those corresponding to our participation in TEPSS project. The discrepancies observed in a 1D simulation of the selected tests will be shown and analyzed. An evaluation of how the 3D version can lead to a better agreement with data will be included. Disadvantages of 3D codes will be shown too. Combining the use of different codes, and considering analyst criteria, will make possible to establish suitable recommendations from both engineering and scientific point of view. (author)

  14. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    The MARS nuclear plant is a 600 MWth PWR with completely passive core safeguards. The most relevant innovative safety system is the Emergency Core Cooling System (ECCS), which is based on natural circulation, and on a passive-type activation that follows a core flow decrease, whatever was the cause (only one component, 400% redundant, is not static). The main thermal hydraulic transients occurring as a consequence of design basis accidents for the MARS plant were presented at the ICONE 3 Conference. Those transients were analyzed in the first stage, with the aim at pointing out the capability of the innovative ECCS to intervene. So, they included only a short-time analysis (extended for a few hundreds of seconds) and the well known RELAP 5 computer program was used for this purpose. In the present paper, the long-term analyses (extended for several thousands of seconds) of the same transients are shown. These analyses confirmed that the performance of the Emergency Core Cooling System of the MARS reactor is guaranteed also in long-term scenarios

  15. Advanced technologies for water cooled reactors 1990. Pt. 2

    International Nuclear Information System (INIS)

    The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  16. Advanced technologies for water cooled reactors 1990. Pt. 1

    International Nuclear Information System (INIS)

    The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

  17. Development of advanced strain diagnostic techniques for reactor environments.

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  18. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  19. Status of advanced light water reactor designs 2004

    International Nuclear Information System (INIS)

    The report is intended to be a source of reference information for interested organizations and individuals. Among them are decision makers of countries considering implementation of nuclear power programmes. Further, the report is addressed to government officials with an appropriate technical background and to research institutes of countries with existing nuclear programmes that wish to be informed on the global status in order to plan their nuclear power programmes including both research and development efforts and means for meeting future. The future utilization of nuclear power worldwide depends primarily on the ability of the nuclear community to further improve the economic competitiveness of nuclear power plants while meeting stringent safety requirements. The IAEA's activities in nuclear power technology development include the preparation of status reports on advanced reactor designs to provide all interested IAEA Member States with balanced and objective information on advances in nuclear plant technology. In the field of light water reactors, the last status report published by the IAEA was 'Status of Advanced Light Water Cooled Reactor Designs: 1996' (IAEA-TECDOC-968). Since its publication, quite a lot has happened: some designs have been taken into commercial operation, others have achieved significant steps toward becoming commercial products, including certification from regulatory authorities, some are in a design optimization phase to reduce capital costs, development for other designs began after 1996, and a few designs are no longer pursued by their promoters. With this general progress in mind, on the advice and with the support of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for Light Water Reactors (LWRs), the IAEA has prepared this new status report on advanced LWR designs that updates IAEA-TECDOC-968, presenting the various advanced LWR designs in a balanced way according to a common outline

  20. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  1. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    International Nuclear Information System (INIS)

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  2. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  3. Application of PSA methodology to design improvement of JAERI Passive Safety Reactor (JPSR)

    International Nuclear Information System (INIS)

    A probabilistic safety assessment (PSA) technique was applied to the design of JAERI Passive Safety Reactor (JPSR). A PSA was performed to clarify safety features and identify vulnerabilities of the original design. Based on the PSA results and considering thermal-hydraulic analyses and experiments, the JPSR design was improved to enhance plant safety. The improved design was re-evaluated with the PSA. Initiating events selected in this study were: large-break LOCA, medium- and small-break LOCAs, SGTR, main steam line break, loss of offsite power, loss of feed water, and other transients. Fault tree analyses were used to evaluate the system unavailabilities. The total core damage frequency due to internal events was estimated to be less than 10-7/RY. The contribution of high frequency non-LOCA events could be significantly reduced by the design modification. The dominant initiating event was the small break LOCA and the dominant sequence was the failure of residual heat removal system. The present study indicated that the improved JPSR design has sufficient safety margin and the PSA methodology is very effective to improve reactor safety systems in a conceptual design phase. (author)

  4. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  5. Advanced Test Reactor National Scientific User Facility Partnerships

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of

  6. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  7. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  8. Advances in Process Intensification through Multifunctional Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    O' Hern, Timothy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center; Evans, Lindsay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Miller, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Cooper, Marcia [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Energetic Components Realization Center; Torczynski, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pena, Donovan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walt [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center

    2011-02-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  9. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed

  10. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    International Nuclear Information System (INIS)

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  11. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  12. Feasibility study of advanced fuel burning nuclear reactors

    International Nuclear Information System (INIS)

    An investigation has been conducted to determine both physics, engineering and economic aspects of fusion power reactors based on magnetic confinement and on burning advanced fuels (AFs). DT burning Tokamaks are taken as reference concept. We show that the attractive features of advanced fuels, in particular of neutronlean proton-based AFs, can be combined, in appropriately designed AF reactors (high beta), with power densities comparable to or even higher than those achievable in DT Tokamaks. Moreover we identify physical requirements which would assure Q values well above unity. As an example a semi-open confinement scheme is analyzed based on a self-consistent plasma calculation. We find that a mirror, even if only ''semi-open'' as a result of strong diamagnetism, can barely be expected to achieve high Q values. Therefore confinement schemes such as compact tori, multipole surmacs etc. may be required to burn AFs. We conclude that the economics of AF reactors, as determined by the nuclear boiler power density, may be superior to that of DT-rectors if low recirculating power fractions can be obtained by appropriate plasma tayloring (high fractional transfer of fusion power to ions required). A more detailed investigation is suggested for proton-based fuel cycles. (orig.)

  13. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  14. The search for advanced remote technology in fast reactor reprocessing

    International Nuclear Information System (INIS)

    Research and development in fast reactor reprocessing has been under way ∼ 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the Experimental Breeder Reactor II (EBR-II) facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. The Federal Republic of Germany (FRG) and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper focuses on the search for improved facility concepts and better maintenance systems in the CFRP, and, in turn, on how developments at ORNL have influenced the technology elsewhere

  15. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellarators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transport behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  16. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellerators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transort behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  17. Balanced Design of Safety Systems of CAREM Advanced Reactor

    International Nuclear Information System (INIS)

    Nuclear Power Plants must meet the performance that the market and the population demand in order to be part of the electricity supply industry.It is related mainly with the results of reactor's economy and safety.New advances in the methodology developed for reactor economic optimization analyzing its safety at an early engineering stage, aiming at balancing these important features of the design, are presented in this work.In particular, the coupling that appears when dimensioning the Emergency Injection System, the Residual Heat Removal System and the containment height of CAREM reactor is described.The new models appended to the computer code that embodies the methodology to balance de designs are shown.Finally the results obtained with the optimizations when applying it are presented.Furthermore, a criterion to establish the maximal diameter for acceptable breaks in RPV's penetrations arises from this work.The application of the methodology and the computer code developed turns out to prove the advantages they provide to reactor design so that the plants are properly balanced and optimized

  18. Design of the reactor vessel inspection robot for the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    A consortium of four universities and Oak Ridge National Laboratory designed a prototype wall-crawling robot to perform weld inspection in an advanced nuclear reactor. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a mock non-hostile environment and shown to perform as expected, as detailed in this report

  19. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  20. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  1. Advanced Space Nuclear Reactors from Fiction to Reality

    Science.gov (United States)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  2. Joining of Oxide Dispersion Strengthened Steels for Advanced Reactors

    Science.gov (United States)

    Baker, B. W.; Brewer, L. N.

    2014-12-01

    The design, manufacture, and experimental analysis of structural materials capable of operation in the high temperatures, corrosive environments, and radiation damage spectra of future reactor designs remain one of the key pacing items for advanced reactor designs. The most promising candidate structural materials are vanadium-based refractory alloys, silicon carbide composites and oxide dispersion strengthened steels. Of these, oxide dispersion strengthened steels are a likely near-term candidate to meet required demands. This paper reviews different variants of oxide dispersion strengthened steels and discusses their capability with regard to high-temperature strength, corrosion resistance, and radiation damage resistance. Additionally, joining of oxide dispersion strengthened steels, which has been cited as a limiting factor preventing their use, is addressed and reviewed. Specifically, friction stir welding of these steels is reviewed as a promising joining method for oxide dispersion strengthened steels.

  3. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  4. Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

    International Nuclear Information System (INIS)

    The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 deg. C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000 deg. C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 deg. C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating

  5. Evolutionary/advanced light water reactor data report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  6. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  7. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    International Nuclear Information System (INIS)

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  8. Metal fires and their implications for advanced reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  9. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  10. Passive core injection system with steam driven jet pump for next generation nuclear reactors

    International Nuclear Information System (INIS)

    The steam driven jet pump (SDJP) is a device without moving parts, in which steam is used as an energy source to pump cold water from a pressure much lower than the steam pressure to a pressure higher than the steam pressure. In this study, a design of a passive core injection system (PCIS) with SDJP is given for application to Boiling Water Reactor (BWR). The operation range of PCIS depends on the operation characteristics of the SDJP. Thus, the operation characteristics of the SDJP has been investigated in terms of independent parameters, such as water temperature, pressure, and steam inlet pressure which is the driving force of the SDJP to obtain a wide operation range. The water tank pressure is chosen as 320 kPa to increase the maximum operation pressure of the SDJP. The supply water temperature is chosen as 15 deg. C which is about room temperature, and the steam which comes from the reactor pressure vessel is assumed as saturated. The operation range of the SDJP in terms of the steam inlet pressure is obtained from 2 to 10.5 MPa. The study shows that when the discharge pressure increases, the discharge mass flow rate and the temperature increases as well. The probabilistic reliability of the PCIS with the SDJP is assessed as well. The reliability assessment of the PCIS with SDJP is investigated using a fault tree, and compared with the core injection system with turbine-driven pump. It is found that the PCIS with SDJP is more reliable than the core injection system with turbine-driven pump

  11. Integration of the functional reliability of two passive safety systems to mitigate a SBLOCA+BO in a CAREM-like reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    Mezio, Federico, E-mail: federico.mezio@cab.cnea.gov.ar [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Grinberg, Mariela [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina); Lorenzo, Gabriel [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Giménez, Marcelo [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina)

    2014-04-01

    Highlights: • An estimation of the Functional Unreliability was performed using RMPS methodology. • The methodology uses an improved response surface in order to estimate the FU. • The FU may become relevant to be analyzed in the Passive Safety Systems. • There were proposed two ways to incorporate the FU into an APS. - Abstract: This paper describes a case study of a methodological approach for assessing the functional reliability of passive safety systems (PSS) and its treatment within a probabilistic safety assessment (PSA). The functional unreliability (FU) can be understood as the failure probability of PSS to fulfill its mission due to the impairment of the related passive safety function. The safety function accomplishment is characterized and quantified by a performance indicator (PI), which is a measure of how far the system is from verifying its mission. PI uncertainties are estimated from uncertainty propagation of selected parameters. A methodology based on the reliability methodology for passive system (RMPS) one is used to estimate the FU associated to the isolation condensers (ICs) in combination with the accumulators (medium pressure injection system) of a CAREM-like integral advanced reactor. A small break loss of coolant accident with black-out is selected as an evaluation case. This implies success of reactor shut-down (inherent) and failure of residual heat removal by active systems. The safety function to accomplish is to refill the reactor pressure vessel (RPV) in order to avoid core damage. For this case, to allow the discharge of accumulators into RPV, the pressure must be reduced by the IC. The methodology for passive safety function assessment considers uncertainties in code parameters, besides uncertainties in engineering parameters (design, construction, operation and maintenance), in order to perform Monte Carlo simulations based on best estimate (B-E) plant model. Then, response surfaces based on PI are used for improving the

  12. On-line interrogation of pebble bed reactor fuel using passive gamma-ray spectrometry

    Science.gov (United States)

    Chen, Jianwei

    The Pebble Bed Reactor (PBR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles (of varying enrichment) are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burnup limit (˜80,000--100,000 MWD/MTU). Unlike the situation with conventional light water reactors (LWRs), depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. In this work, an investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line interrogation of PBR fuel for the simultaneous determination of burnup and enrichment on a pebble-by-pebble basis. Due to the unavailability of irradiated or fresh pebbles, Monte Carlo simulations were used to study the gamma-ray spectra of the PBR fuel at various levels of burnup. A pebble depletion calculation was performed using the ORIGEN code, which yielded the gamma-ray source term that was introduced into the input of an MCNP simulation. The MCNP simulation assumed the use of a high-purity coaxial germanium detector. Due to the lack of one-group high temperature reactor cross sections for ORIGEN, a heterogeneous MCNP model was developed to describe a typical PBR core. Subsequently, the code MONTEBURNS was used to couple the MCNP model and ORIGEN. This approach allowed the development of the burnup-dependent, one-group spectral-averaged PBR cross sections to be used in the ORIGEN pebble depletion calculation. Based on the above studies, a relative approach for performing the measurements was established. The approach is based on using the relative activities of Np-239/I-132 in combination

  13. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h)-1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  14. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  15. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  16. Passive and Active Fast-Neutron Imaging in Support of Advanced Fuel Cycle Initiative Safeguards Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Blackston, Matthew A [ORNL; Hausladen, Paul [ORNL

    2010-04-01

    Results from safeguards-related passive and active coded-aperture fast-neutron imaging measurements of plutonium and highly enriched uranium (HEU) material configurations performed at Idaho National Laboratory s Zero Power Physics Reactor facility are presented. The imaging measurements indicate that it is feasible to use fast neutron imaging in a variety of safeguards-related tasks, such as monitoring storage, evaluating holdup deposits in situ, or identifying individual leached hulls still containing fuel. The present work also presents the first demonstration of imaging of differential die away fast neutrons.

  17. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  18. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  19. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  20. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  1. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Science.gov (United States)

    Wali, W. A.; Hassan, K. H.; Cullen, J. D.; Al-Shamma'a, A. I.; Shaw, A.; Wylie, S. R.

    2011-08-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  2. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    International Nuclear Information System (INIS)

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  3. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wali, W A; Hassan, K H; Cullen, J D; Al-Shamma' a, A I; Shaw, A; Wylie, S R, E-mail: w.wali@2009.ljmu.ac.uk [Built Environment and Sustainable Technologies Institute (BEST), School of the Built Environment, Faculty of Technology and Environment Liverpool John Moores University, Byrom Street, Liverpool L3 3AF (United Kingdom)

    2011-08-17

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  4. Economics of an advanced pressurized water reactor (APWR)

    International Nuclear Information System (INIS)

    Limited natural uranium resources together with their low utilization in current lightwater reactors (LWR) on the one hand and the high capital investments for a LWR and fast breeder reactor system resulting in a high fuel utilization are the most important reasons for research and development (R+D) work related to a high converting APWR system. It is the main task of this analysis to study the economic behaviour of an APWR combining its technical and physical parameters with an economic data base. After introductional remarks chapter II presents the potential improvements of the uranium utilization and their importance for a whole national economy. Chapter III shows the micro-economic aspects for an electricity producing utility using such an APWR plant. The chapter is restricted to the fuel cycle costs and their dependence on various parameters. The corresponding costs of other nuclear power plants are described in chapter IV and compared to those of the APWR in chapter V. Finally a cost comparison on the basis of the electricity generating costs will complete the economic picture of an advanced pressurized water reactor. (orig./UA)

  5. The Advanced Test Reactor as a National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) has been in operation since 1967 and mainly used to support U.S. Department of Energy (US DOE) materials and fuels research programs. Irradiation capabilities of the ATR and post-irradiation examination capabilities of the Idaho National Laboratory (INL) were generally not being utilized by universities and other potential users due largely to a prohibitive pricing structure. While materials and fuels testing programs using the ATR continue to be needed for US DOE programs such as the Advanced Fuel Cycle Initiative and Next Generation Nuclear Plant, US DOE recognized there was a national need to make these capabilities available to a broader user base. In April 2007, the U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). As a NSUF, most of the services associated with university experiment irradiation and post-irradiation examinations are provided free-of-charge. The US DOE is providing these services to support U.S. leadership in nuclear science, technology, and education and to encourage active university/industry/laboratory collaboration. The first full year of implementing the user facility concept was 2008 and it was a very successful year. The first university experiment pilot project was developed in collaboration with the University of Wisconsin and began irradiation in the ATR in 2008. Lessons learned from this pilot program will be applied to future NSUF projects. Five other university experiments were also competitively selected in March 2008 from the initial solicitation for proposals. The NSUF now has a continually open process where universities can submit proposals as they are ready. Plans are to invest in new and upgraded capabilities at the ATR, post-irradiation examination capabilities at the INL, and in a new experiment assembly facility to further support the implementation of the user facility concept. Through a newly created Partnership Program

  6. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

  7. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  8. Reliability assurance programme guidebook for advanced light water reactors

    International Nuclear Information System (INIS)

    To facilitate the implementation of reliability assurance programmes (RAP) within future advanced reactor programmes and to ensure that the next generation of commercial nuclear reactors achieves the very high levels of safety, reliability and economy which are expected of them, in 1996, the International Atomic Energy Agency (IAEA) established a task to develop a guidebook for reliability assurance programmes. The draft RAP guidebook was prepared by an expert consultant and was reviewed/modified at an Advisory Group meeting (7-10 April 1997) and at a consults meeting (7-10 October 1997). The programme for the RAP guidebook was reported to and guided by the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR). This guidebook will demonstrate how the designers and operators of future commercial nuclear plants can exploit the risk, reliability and availability engineering methods and techniques developed over the past two decades to augment existing design and operational nuclear plant decision-making capabilities. This guidebook is intended to provide the necessary understanding, insights and examples of RAP management systems and processes from which a future user can derive his own plant specific reliability assurance programmes. The RAP guidebook is intended to augment, not replace, specific reliability assurance requirements defined by the utility requirements documents and by individual nuclear steam supply system (NSSS) designers. This guidebook draws from utility experience gained during implementation of reliability and availability improvement and risk based management programmes to provide both written and diagrammatic 'how to' guidance which can be followed to assure conformance with the specific requirements outlined by utility requirements documents and in the development of a practical and effective plant specific RAP in any IAEA Member State

  9. An experimental substantiation of the design functions imposed on the additional system for passively flooding the core of a VVER reactor

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from a research work on experimentally substantiating the serviceability of the additional system for passively flooding the core of a VVER reactor from the second-stage hydro accumulators are presented.

  10. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  11. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 oC which is substantially lower than ∼627 oC as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  12. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hung, T.C., E-mail: tchung@ntut.edu.t [Department of Mechanical Engineering, National Taipei University of Technology, 1, Sec. 3, Chung-hsiao E. Rd., Taipei 10608, Taiwan (China); Dhir, V.K. [Department of Mechanical and Aerospace Engineering, UCLA, CA (United States); Chang, J.C. [Graduate Institute of Mechanical and Electrical Engineering, National Taipei University of Technology, Taiwan (China); Wang, S.K. [Department of Mechanical and Automation Engineering, I-Shou University, Taiwan (China)

    2011-01-15

    Research highlights: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW) The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to {approx}551 {sup o}C which is substantially lower than {approx}627 {sup o}C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity

  13. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  14. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  15. Supervisory Control System Architecture for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cetiner, Sacit M [ORNL; Cole, Daniel L [University of Pittsburgh; Fugate, David L [ORNL; Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Muhlheim, Michael David [ORNL; Rao, Nageswara S [ORNL; Wood, Richard Thomas [ORNL

    2013-08-01

    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  16. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  17. The neutron texture diffractometer at the China Advanced Research Reactor

    Science.gov (United States)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  18. Advanced nuclear reactor public opinion project. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  19. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  20. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  1. Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

  2. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  3. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    OpenAIRE

    Kulesza Joel A.; Franceschini Fausto; Evans Thomas M.; Gehin Jess C.

    2016-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a fir...

  4. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  5. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  6. Modelling of advanced structural materials for GEN IV reactors

    Science.gov (United States)

    Samaras, M.; Hoffelner, W.; Victoria, M.

    2007-09-01

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors.

  7. Modelling of advanced structural materials for GEN IV reactors

    International Nuclear Information System (INIS)

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors

  8. Advances in passive-remote and extractive Fourier transform infrared spectroscopic systems

    International Nuclear Information System (INIS)

    The Clean Air Act of 1990 requires the monitoring of air toxics including those from incinerator emissions. Continuous emission monitors (CEM) would demonstrate the safety of incinerators and address public concern about emissions of hazardous organic compounds. Fourier transform infrared (FTIR) spectroscopy can provide the technology for continuous emission monitoring of stacks. Stack effluent can be extracted and analyzed in less than one minute with conventional FTIR spectrometers. Passive-remote FTIR spectrometers can detect certain emission gases over 1 km away from a stack. The authors discuss advances in both extractive and passive-remote FTIR technology. Extractive systems are being tested with EPA protocols, which will soon replace periodic testing methods. Standard operating procedures for extractive systems are being developed and tested. Passive-remote FTIR spectrometers have the advantage of not requiring an extracted sample; however, they have less sensitivity. We have evaluated the ability of commercially available systems to detect fugitive plumes and to monitor carbon monoxide at a coal-fired power plant

  9. Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

    International Nuclear Information System (INIS)

    The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

  10. BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

    International Nuclear Information System (INIS)

    As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

  11. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  12. The advanced test reactor national scientific user facility: advancing nuclear technology education

    International Nuclear Information System (INIS)

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy designated the Idaho National Laboratory (INL) Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The ATR NSUF provides education programs including a Users Week, internships, faculty student team projects and faculty/staff exchanges. In addition, the ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  13. Advanced CANDU reactor development: a customer-driven program

    International Nuclear Information System (INIS)

    The Advanced CANDU Reactor (ACR) product development program is well under way. The development approach for the ACR is to ensure that all activities supporting readiness for the first ACR project are carded out in parallel, as parts of an integrated whole. In this way design engineering, licensing, development and testing, supply chain planning, construct ability and module strategy, and planning for commissioning and operations, all work in synergy with one another. Careful schedule management :ensures that program focus stays on critical path priorities.'This paper provides an overview of the program, with an emphasis on integration to ensure maximum project readiness, This program management approach is important now that AECL is participating as the reactor vendor in Dominion Energy's DOE-sponsored Combined Construction/Operating License (COL) program. Dominion Energy selected the ACR-700 as their reference reactor technology for purposes of demonstrating the COL process. AECL's development of the ACR is unique in that pre-licensing activities are being carded out parallel in the USA and Canada, via independent, but well-communicated programs. In the short term, these programs are major drivers of ACR development. The ACR design approach has been to optimize to achieve major design objectives: capital cost reduction, robust design with ample margins, proveness by using evolutionary change from existing :reference plants, design for ease :of operability. The ACR development program maintains these design objectives for each of the program elements: Design: .Carefully selected design innovations based on the SEU fuel/light water coolant:/heavy water moderator approach. Emphasis on lessons-learned review from operating experience and customer feedback Licensing: .Safety case based on strengths of existing CANDU plus benefits of optimised design Development and Test: Choice of materials, conditions to enable incremental testing building on existing CANDU and LWR

  14. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  15. Advances in shielding calculations for the PEC reactor

    International Nuclear Information System (INIS)

    In this paper calculations of neutron and gamma streaming through various penetrations in the plug and neutron shield of the sodium cooled fast reactor PEC, currently under construction, are described. The object of the calculations has been to verify the accessibility, 3 days after reactor shut-down, of the area directly above the reactor. (author)

  16. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  17. Advances in crack-arrest technology for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs.

  18. On-Line NDE for Advanced Reactor Designs

    Science.gov (United States)

    Nakagawa, N.; Inanc, F.; Thompson, R. B.; Junker, W. R.; Ruddy, F. H.; Beatty, J. M.; Arlia, N. G.

    2003-03-01

    This expository paper introduces the concept of on-line sensor methodologies for monitoring the integrity of components in next generation power systems, and explains general benefits of the approach, while describing early conceptual developments of suitable NDE methodologies. The paper first explains the philosophy behind this approach (i.e. the design-for-inspectability concept). Specifically, we describe where and how decades of accumulated knowledge and experience in nuclear power system maintenance are utilized in Generation IV power system designs, as the designs are being actively developed, in order to advance their safety and economy. Second, we explain that Generation IV reactor design features call for the replacement of the current outage-based maintenance by on-line inspection and monitoring. Third, the model-based approach toward design and performance optimization of on-line sensor systems, using electromagnetic, ultrasonic, and radiation detectors, will be explained. Fourth, general types of NDE inspections that are considered amenable to on-line health monitoring will be listed. Fifth, we will describe specific modeling developments to be used for radiography, EMAT UT, and EC detector design studies.

  19. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, fast breeder reactors are to be developed as the main of the future nuclear power generation in Japan, and when their development is advanced, it has been decided to positively aim at building up the plutonium utilization system using FBRs superior to the uranium utilization system using LWRs. Also it has been decided that the development of FBRs requires to exert incessant efforts for a considerable long period under the proper cooperation system of government and people, and as for its concrete development, hereafter the deliberation is to be carried out in succession by the expert subcommittee on FBR development projects of the Atomic Energy Commission. The subcommittee was founded in May, 1986, to deliberate on the long term promotion measures for FBR development, the measures for promoting the research and development, the examination of the basic specification of a demonstration FBR, the measures for promoting international cooperation, and other important matters. As the results of investigation, the situation around the development of FBRs, the fundamentals at the time of promoting the research and development, the subjects of the research and development and so on are reported. (Kako, I.)

  20. AP1000R pressurised water reactor project in china advances toward completion

    International Nuclear Information System (INIS)

    The AP1000R pressurised water reactor (PWR) project in China is the first deployment of its first-of-a-kind Generation III+ technology, making it one of most internationally important and industry-significant new build projects. The innovative AP1000 PWR design contains advanced passive safety and performance features that involve fewer active safety components than a traditional plant, thereby reducing the site footprint. The AP1000 reactor is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission, and it has received an Interim Design Acceptance Confirmation from the Office for Nuclear Regulation and an Interim Statement of Design Acceptability from the Environment Agency in the United Kingdom. Construction and testing of dual AP1000 PWR units is currently in progress in each of two coastal sites in the People's Republic of China: Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Since the initial contract award in 2007, the Westinghouse Consortium has worked in concert with the owners to construct the plants using innovative structural and mechanical modules. Uniquely designed plant components and essential instrumentation and control systems have been manufactured, delivered, and installed at the plants. Numerous personnel, including future reactor operators, have been trained at both the Sanmen and Haiyang sites, and technology transfer of technical documents and computer codes is well underway. The commercial operation dates are now nearing for Sanmen Unit 1 and Haiyang Unit 1, the first two units scheduled for completion. Consequently, these units are now in advanced stages of completion and present activities include planning and preparation for pre-operational testing, system turnover, and commissioning leading to fuel load, and eventual commercial operation. These activities are pioneering, in that they have never before been performed for a new build of

  1. Thermohydraulic relationships for advanced water cooled reactors, and the role of IAEA

    International Nuclear Information System (INIS)

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1999. It was included into the IAEA's Programme following endorsement in 1995 by the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote international information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. (authors)

  2. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    Science.gov (United States)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  3. The IRIS consortium: international cooperation in advanced reactor development

    International Nuclear Information System (INIS)

    Besides its many outstanding technical innovations in the design and safety, the most innovative feature of the International Reactor Innovative and Secure (IRIS), is perhaps the international cooperation which carries on its development. IRIS is designed by an international consortium which currently numbers 21 organizations from ten countries across four continents. It includes reactor, fuel and fuel cycle vendors, component manufacturers, laboratories, academia, architect engineers and power producers. The defining organizational characteristics of IRIS is that while Westinghouse has overall lead and responsibility, this lead is of the type of 'primus inter pares' (first among equals) rather than the traditional owner versus suppliers/contractors relationship. All members of the IRIS consortium contribute and expect to have a return, should IRIS be successfully deployed, commensurate to their investment. The nature of such return will be tailored to the type of each organization, because it will of course be of a different nature for say a component manufacturer, university, or architect engineer. One fundamental tenet of the consortium is that all members, regardless of their amount of contribution, have equal access to all information developed within the project. Technical work is thus being coordinated by integrated subgroups and the whole team meets twice a year to perform an overall review of the work, discuss policy and strategy and plan future activities. Personnel from consortium members have performed internships, mostly at Westinghouse locations in Pittsburgh, Pennsylvania, and Windsor, Connecticut, but also at other members, as it has been the case for several graduate students. In fact, more than one hundred students at the various universities have been working on IRIS, most of them conducting graduate theses at the master or doctoral level. The IRIS experience has proved very helpful to the students in successfully landing their employment choice

  4. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  5. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    Energy Technology Data Exchange (ETDEWEB)

    T. R. Allen; J. B. Benson; J. A. Foster; F. M. Marshall; M. K. Meyer; M. C. Thelen

    2009-05-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  6. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  7. Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis

    International Nuclear Information System (INIS)

    This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multiphysics simulation tools that leverage high performance computing to reduce uncertainty in the prediction of the performance and safety of advanced reactor and fuel designs. The toolkit effort is composed of two major components, the fuels product line, which provides tools for fuel performance analysis, and the reactor product line, which provides tools for reactor performance and safety analysis. This paper presents an overview of the NEAMS reactor product line development effort. (author)

  8. Deterministic and risk-informed approaches for safety analysis of advanced reactors: Part I, deterministic approaches

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang Kyu [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-05-15

    The objective of this paper and a companion paper in this issue (part II, risk-informed approaches) is to derive technical insights from a critical review of deterministic and risk-informed safety analysis approaches that have been applied to develop licensing requirements for water-cooled reactors, or proposed for safety verification of the advanced reactor design. To this end, a review was made of a number of safety analysis approaches including those specified in regulatory guides and industry standards, as well as novel methodologies proposed for licensing of advanced reactors. This paper and the companion paper present the review insights on the deterministic and risk-informed safety analysis approaches, respectively. These insights could be used in making a safety case or developing a new licensing review infrastructure for advanced reactors including Generation IV reactors.

  9. Metal removal efficiency and ecotoxicological assessment of field-scale passive treatment biochemical reactors

    Science.gov (United States)

    Anaerobic biochemical reactors (BCRs) are useful for removing metals from mining-impacted water (MIW) at remote sites. Removal processes include sorption and precipitation of metal sulfides, carbonates and hydroxides. A question of interest is whether BCRs remove aquatic toxicit...

  10. Practical application of passive safety features for the advanced neutron source cooling system

    International Nuclear Information System (INIS)

    The results of a conceptual design study leading to the definition of a reference design for the Advanced Neutron Source (ANS) heavy water cooling system are presented. The objective of this study was to define a cooling system that not only met the ANS goals for operating parameters, reliability, availability, and maintainability, but also used inherent, passive, and diverse features and characteristics to satisfy the ANS internal events core melt goal of -5/yr. The approach taken in this study was to define a cooling system configuration having the minimum basic components and characteristics to satisfy the requirements for normal operation, and then to add only those features necessary to meet the requirements for all emergency design-basis events

  11. Jitter Suppression Via Reaction Wheel Passive Isolation for the NASA Advanced X-Ray Astrophysics Facility

    Science.gov (United States)

    Pendergast, Karl J.; Schauwecker, Chris J.

    1998-01-01

    Text: Third in the series of NASA great observatories, the Advanced X-ray Astrophysics Facility (AXAF) is scheduled for launch from the Space Shuttle in September 1998. Following in the path of the Hubble Space Telescope and the Compton Gamma Ray Observatory, this telescope will image light at x-ray wavelengths, facilitating the detailed study of such phenomena as supernovae and quasars. The AXAF program is sponsored by the Marshall Space Flight Center (MSFC) in Huntsville, Alabama. Due to exacting requirements on the performance of the AXAF optical system, it is necessary to reduce the transmission of reaction wheel jitter disturbances to the observatory. This reduction is accomplished via use of a passive mechanical isolation system which acts as an interface between the reaction wheels and the spacecraft central structure.

  12. Mirror Advanced Reactor Study (MARS): executive summary and overview

    International Nuclear Information System (INIS)

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (2), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li17Pb83) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (10000C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter

  13. Mirror Advanced Reactor Study (MARS): executive summary and overview

    Energy Technology Data Exchange (ETDEWEB)

    Logan, B.G.; Perkins, L.J.; Gordon, J.D.

    1984-07-01

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (<10 W/cm/sup 2/), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li/sub 17/Pb/sub 83/) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (1000/sup 0/C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter.

  14. Design features of Advanced Power Reactor (APR) 1400 steam generator

    International Nuclear Information System (INIS)

    Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80+, which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80+ steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80+. Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

  15. Design of GA thermochemical water-splitting process for the Mirror Advanced Reactor System

    International Nuclear Information System (INIS)

    GA interfaced the sulfur-iodine thermochemical water-splitting cycle to the Mirror Advanced Reactor System (MARS). The results of this effort follow as one section and part of a second section to be included in the MARS final report. This section describes the process and its interface to the reactor. The capital and operating costs for the hydrogen plant are described

  16. Advanced Reactor Safety Research Program quarterly report, April--June 1977. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-11-01

    Information is presented concerning accident energetics; core debris behavior; sodium containment and structural integrity; research for elevated temperature design criteria; fuel motion detection; ACPR fuel motion system; and advanced reactor safety research assessment.

  17. Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

    2015-05-15

    As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

  18. Development of advanced PWR system analysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Y. D.; Kim, S. O.; Jung, B. D.; Kim, Y. I.; Chang, M. H.; Lee, Y. J.; Yun, J. H.

    1997-12-31

    The scope of this project is to establish the basic analysis technologies for the advanced designed with the passive and inherent safety concepts. The scope is extended to the application of these technologies to the performance and safety analysis of the passive reactor. Since the different design concepts are applied depending on the reactor power, the study is conducted for the small and medium sized integral reactor as well as the large scale passive reactors by focusing on the analysis technology development for the passive components. The design concepts which can be applied for the safety enhancement of the domestic advanced reactor are developed through evaluating the technical information of the overseas advanced reactor concepts.

  19. Study on transient hydrogen behavior and effect on passive containment cooling system of the advanced PWR

    International Nuclear Information System (INIS)

    A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)

  20. Nuclear developments: the DMAX advanced reactor control system

    International Nuclear Information System (INIS)

    Framatome has recently developed a new system for controlling the rod cluster control assemblies of pressurized water reactors, called the DMAX. The associated reactor control method is called 'mode X'. The DMAX system will be installed in all 'N4' model Framatome nuclear steam supply systems, the first two of which are presently under construction on the Chooz site in France. It will enable fine controlling of the reactor coolant temperature and the axial power offset, entirely automatically, due to double closed-loop regulation. The new DMAX system allows temperature control and continuous maintenance of a stable reactor core power distribution, because of an original method for controlling the movements of the control rods within the reactor. The disturbing xenon oscillations are practically eliminated and the operator is freed from the need of constantly monitoring the axial power offset, which is necessary in the commonly used 'A' or 'G' control modes. The probability of penalizing initial conditions in case an incident or accident occurs is considerably reduced in mode X, with the DMAX system, and the reactor's load-following performances are improved. In addition, the reactivity variations that must necessarily be compensated for in mode G by changing the boric acid concentration of the reactor coolant can be simply compensated for by control rod movements in mode X. This possibility yields a major reduction in the volume of liquid effluents that must subsequently be created. The system is outlined and its operation explained. (author)

  1. Physics-Based Multi-State Models of Passive Component Degradation for the R7 Reactor Simulation Environment

    Energy Technology Data Exchange (ETDEWEB)

    Unwin, Stephen D.; Layton, Robert F.; Johnson, Kenneth I.; Lowry, Peter P.

    2012-06-25

    Abstract: The Next Generation Systems Analysis Code - referred to as R7 - is reactor systems simulation software being developed to support the Risk-Informed Safety Margin Characterization Pathway of the U.S. Department of Energy's Light Water Reactor Sustainability Program. It will provide an integrated multi-physics environment, implemented in an uncertainty quantification (UQ) framework that can produce risk and other performance insights on long-term reactor operations. An element of this simulation environment will be the performance of passive components and materials. Conventional models of component reliability are largely parametric, relying on plant service data to estimate component lifetimes and failure rates. This type of model has limited usefulness in the R7 environment where the intent is to explicitly determine the influence of physical stressors on component degradation. In this paper, we describe a new class of multi-state physics-based component models designed to be R7-compatible. These models capture the physics of materials degradation while also incorporating the effects of interventions and component rejuvenation. The models are implemented in a cumulative damage framework that allows the impact of an evolving physical environment to be addressed without recourse to resampling within the Monte Carlo-based UQ framework. The paper describes an application to stress corrosion cracking in dissimilar metal welds - a principal contributor to potential loss of coolant accidents. So while R7 will have the more conventional capability of reactor simulation codes to model the impact of degraded components and systems on plant performance, the methodology described here allows R7 to model the inverse effect; the impact of the physical environment on component degradation and performance.

  2. Power maximization method for land-transportable fully passive lead–bismuth cooled small modular reactor systems

    International Nuclear Information System (INIS)

    Highlights: • The power maximization method for LBE natural circulation cooled SMRs was developed. • The two powers in view of neutronics and thermal-hydraulics were considered. • The limitations for designing of LBE natural circulation cooled SMRs were summarized. • The necessary conditions for safety shutdown in accidents were developed. • The maximized power in the case study is 206 MW thermal. - Abstract: Although current pressurized water reactors (PWRs) have significantly contributed to global energy supply, PWR technology has not been considered a trustworthy energy solution owing to its problems of spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, a lead–bismuth eutectic (LBE) fully passive cooling small modular reactor (SMR) system is suggested. This technology can not only provide the solution for the problems of SNFs through the transmutation feature of the LBE coolant, but also strengthen safety and economy through the concept of natural circulation cooling SMRs. It is necessary to maximize the advantages, namely safety and economy, of this type of nuclear power plants for broader applications in the future. Accordingly, the objective of this study is to maximize the reactor core power while satisfying the limitations of shipping size, materials endurance, and criticality of a long-burning core as well as safety under beyond design basis events. To achieve these objectives, the design limitations of natural circulating LBE-cooling SMRs are derived. Then, the power maximization method is developed based on obtaining the design limitations. The results of this study are expected to contribute to the effectiveness of the reactor design stage by providing insights to designers, as well as by formulating methods for the power maximization of other types of SMRs

  3. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  4. A concept of an advanced inertia fusion reactor; TAKANAWA-I

    International Nuclear Information System (INIS)

    A concept of an advanced inertia fusion reactor: TAKANAWA-I is proposed. A pellet with DT ignitor and DD major fuel, Pb wet walls, C or SiC blocks for shielding, and SiC vessels in the water pool are employed. This reactor does not need blanckets for T breeding, since T is supplied through DD reaction, and has low induced radioactivities. These and a simple structure might give a hopeful prediction of economical and safe advantages and mitigate difficulties of reactor technologies, especially remote maintenance of the reactor. (author)

  5. Inherent safety of advanced nuclear engineering based on BN-800 - type fast reactors

    International Nuclear Information System (INIS)

    Considerations based on the prolonged experience of fast reactor operations exhibiting outlook application of reactors on a basis of BN-800 with sodium coolant are given. Reliability and safety of the block are supported by the probability analysis of safety in the content of engineering project. Conversion on the reactor core with nitride fuel will significantly raise a possibility to conform to safety and nonproliferation of fission materials needs. The suggested optimum variant for reactor core on a basis of nitride fuel is advanced

  6. Numerical Simulation of Flow Field in Flow-guide Tank of China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The flow-guide tank of China advanced research reactor (CARR) is located at the top of the reactor vessel and connected with the inlet coolant pipe. It acts as a reactor inlet coolant distributor and plays an important role in reducing the flow-induced vibration of the internal components of the reactor core. Several designs of the flow-guide tank have been proposed, however, the final design option has to be made after detailed investigation of the velocity profile within the flow-guide tank for each configuration.

  7. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    International Nuclear Information System (INIS)

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

  8. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  9. Development of an advanced antineutrino detector for reactor monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Classen, T., E-mail: classen2@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bernstein, A.; Bowden, N.S. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Cabrera-Palmer, B. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Ho, A.; Jonkmans, G. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada); Kogler, L.; Reyna, D. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Sur, B. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-01-21

    Here we present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. This paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass per detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.

  10. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  11. Advanced-power-reactor design concepts and performance characteristics

    Science.gov (United States)

    Davison, H. W.; Kirchgessner, T. A.; Springborn, R. H.; Yacobucci, H. G.

    1974-01-01

    Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

  12. On the interest of carbon-coated plasma reactor for advanced gate stack etching processes

    International Nuclear Information System (INIS)

    In integrated circuit fabrication the most wide spread strategy to achieve acceptable wafer-to-wafer reproducibility of the gate stack etching process is to dry-clean the plasma reactor walls between each wafer processed. However, inherent exposure of the reactor walls to fluorine-based plasma leads to formation and accumulation of nonvolatile fluoride residues (such as AlFx) on reactor wall surfaces, which in turn leads to process drifts and metallic contamination of wafers. To prevent this while keeping an Al2O3 reactor wall material, a coating strategy must be used, in which the reactor is coated by a protective layer between wafers. It was shown recently that deposition of carbon-rich coating on the reactor walls allows improvements of process reproducibility and reactor wall protection. The authors show that this strategy results in a higher ion-to-neutral flux ratio to the wafer when compared to other strategies (clean or SiOClx-coated reactors) because the carbon walls load reactive radical densities while keeping the same ion current. As a result, the etching rates are generally smaller in a carbon-coated reactor, but a highly anisotropic etching profile can be achieved in silicon and metal gates, whose etching is strongly ion assisted. Furthermore, thanks to the low density of Cl atoms in the carbon-coated reactor, silicon etching can be achieved almost without sidewall passivation layers, allowing fine critical dimension control to be achieved. In addition, it is shown that although the O atom density is also smaller in the carbon-coated reactor, the selectivity toward ultrathin gate oxides is not reduced dramatically. Furthermore, during metal gate etching over high-k dielectric, the low level of parasitic oxygen in the carbon-coated reactor also allows one to minimize bulk silicon reoxidation through HfO2 high-k gate dielectric. It is then shown that the BCl3 etching process of the HfO2 high-k material is highly selective toward the substrate in the

  13. Turning points in reactor design

    International Nuclear Information System (INIS)

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

  14. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  15. Consideration of severe accidents in design of advanced WWER reactors

    International Nuclear Information System (INIS)

    Severe accident related requirements formulated in General Regulations for Nuclear Power Plant Safety (OPB-88), in Nuclear Safety Regulations for Nuclear Power Stations' Reactor Plants (PBYa RU AS-89) and in other NPP nuclear and radiation guides of the Russian Gosatomnadzor are analyzed. In accordance with these guides analyses of beyond design basis accidents should be performed in the reactor plant design. Categorization of beyond design basis accidents leading to severe accidents should be made on occurrence probability and severity of consequences. Engineered features and measures intended for severe accident management should be provided in reactor plant design. Requirements for severe accident analyses and for development of measures for severe accident management are determined. Design philosophy and proposed engineered measures for mitigation of severe accidents and decrease of radiation releases are demonstrated using examples of large, WWER-1000 (V-392), and medium size WWER-640 (V-407) reactor plant designs. Mitigation of severe accidents and decrease of radiation releases are supposed to be conducted on basis of consistent realization of the defense in depth concept relating to application of a system of barriers on the path of spreading of ionizing radiation and radioactive materials to the environment and a set of engineered measures protecting these barriers and retaining their effectiveness. Status of fulfilled by OKB Gidropress and other Russian organizations experimental and analytical investigations of severe accident phenomena supporting design decisions and severe accident management procedures is described. Status of the works on retention of core melt inside the WWER-640 reactor vessel is also characterized

  16. Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

  17. An evolutionary approach to advanced water cooled reactors

    International Nuclear Information System (INIS)

    Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP's are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR's/A-PHWR's and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complex machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative assessment of integral reactor concept. (author)

  18. Advanced in the neutron feedback ICF reactor concept

    International Nuclear Information System (INIS)

    Results are reviewed and updated from an earlier design study of a novel nuclear-pumped flashlamp laser (NP-FL) inertial fusion energy (IFE) power reactor based on the neutron feedback concept for IFE. This concept includes nuclear pumping of the laser flashlamp, a D-T seeded D-3He target and magnetic protection of the first wall of the reactor chamber coupled with direct conversion of deflected charged particles. Advantages include an increased overall plant efficiency due to improved energy coupling via neutron feedback, increased thermal-to-electric energy conversion efficiency, and lower neutron activation and waste. These factors are reflected in a driver energy of 5 MJ and a target gain of only 50 for a 53 % efficient 1000-MWe power plant operating at 6 Hz, novel components involved. However, they require further technological development. Consequently, the NP-FL plant appears to provide a very attractive 'second-generation' IFE reactor. (authors)

  19. Advances in zirconium technology for nuclear reactor application

    International Nuclear Information System (INIS)

    Zirconium alloys are extensively used as a material for cladding nuclear fuels and for making core structurals of water-cooled nuclear power reactors all over the world for generation of nearly 16 percent of the worlds electricity. Only four countries in the world, namely France, USA, Russia and India, have large zirconium industry and capability to manufacture reactor grade zirconium sponge, a number of zirconium alloys and a wide variety of structural components for water cooled nuclear reactor. The present paper summarises the status of zirconium technology and highlights the achievement of Nuclear Fuel Complex during the last ten years in developing a wide variety of zirconium alloys and components for water-cooled nuclear power programme

  20. Recent advances on polymeric membranes for membrane reactors

    KAUST Repository

    Buonomenna, M. G.

    2012-06-24

    Membrane reactors are generally applied in high temperature reactions (>400 °C). In the field of fine chemical synthesis, however, much milder conditions are generally applicable and polymeric membranes were applied without their damage. The successful use of membranes in membrane reactors is primary the result of two developments concerning: (i) membrane materials and (ii) membrane structures. The selection of a suited material and preparation technique depends on the application the membrane is to be used in. In this chapter a review of up to date literature about polymers and configuration catalyst/ membranes used in some recent polymeric membrane reactors is given. The new emerging concept of polymeric microcapsules as catalytic microreactors has been proposed. © 2012 Bentham Science Publishers. All rights reserved.

  1. Advances in process intensification through multifunctional reactor engineering.

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, Marcia A.; Miller, James Edward; O' Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  2. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available The Consortium for Advanced Simulation of Light Water Reactors (CASL was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major “Challenge Problems” in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  3. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    Science.gov (United States)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  4. Integral design concepts of advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Under the sub-programme on non-electrical applications of advanced reactors, the International Atomic Energy Agency has been providing a worldwide forum for exchange of information on integral reactor concepts. Two Technical Committee meetings were held in 1994 and 1995 on the subject where state-of-the-art developments were presented. Efforts are continuing for the development of advanced nuclear reactors of both evolutionary and innovative design, for electricity, co-generation and heat applications. While single purpose reactors for electricity generation may require small and medium sizes under certain conditions, reactors for heat applications and co-generation would be necessary in the small and medium range and need to be located closer to the load centres. The integral design approach to the development of advanced light water reactors has received special attention over the past few years. Several designs are in the detailed design stage, some are under construction, one prototype is in operation. A need has been felt for guidance on a number of issues, ranging from design objectives to the assessment methodology needed to show how integral designs can meet these objectives, and also to identify their advantages and problem areas. The technical document addresses the current status of the design, safety and operational issues of integral reactors and recommends areas for future development

  5. The Microbial Community of a Passive Biochemical Reactor Treating Arsenic, Zinc, and Sulfate-Rich Seepage

    OpenAIRE

    Baldwin, Susan Anne; Khoshnoodi, Maryam; Rezadehbashi, Maryam; Taupp, Marcus; Hallam, Steven; Mattes, Al; Sanei, Hamed

    2015-01-01

    Sulfidogenic biochemical reactors (BCRs) for metal removal that use complex organic carbon have been shown to be effective in laboratory studies, but their performance in the field is highly variable. Successful operation depends on the types of microorganisms supported by the organic matrix, and factors affecting the community composition are unknown. A molecular survey of a field-based BCR that had been removing zinc and arsenic for over 6 years revealed that the microbial community was dom...

  6. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  7. A study on the Economic Survey and Analysis for Seawater Desalination Plant using Advanced Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae Chun; Oh, Se Gi; Yi, Jae Kyung; Park, Kyung Seok; Lee, In Gu [Institute for Advanced Engineering, Yongin (Korea, Republic of)

    1997-05-01

    Current status of nuclear desalination plant developments, design characteristics of nuclear reactors and desalination plants, which comprise the nuclear desalination plants, have been reviewed. Based on this review, survey on the possible coupling options has been performed and basic economic evaluation methodology has been investigated. The economic evaluation of the nuclear seawater desalination plant using Advanced Integral Reactor has been performed using the IAEA economic evaluation model for nuclear desalination plants with respect to nuclear-desalination options. The economic cost data of the Advanced Integral Reactor which are required by the IAEA model have been inferred from previous technical data and experiences related with economic evaluation of nuclear power plants by applying appropriate approximation and assumption considering characteristics of advanced Integral Reactor. And the given experienced values have been used for the desalination plant data. The results have been compared with the economic results for other desalination facilities using other types of nuclear reactors and energy sources. Also, sensitivity analyses have been made for major input parameters and variables. The evaluation results show that the seawater desalination plant using Advanced Integral Reactor has economic feasibility compared with desalination plant with other energy sources. 32 refs., 13 tabs., 25 figs. (author)

  8. Commercial scale performance predictions for high-temperature electrolysis plants coupled to three advanced reactor types

    International Nuclear Information System (INIS)

    This paper presents results of system analyses that have been developed to assess the hydrogen-production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor - power-cycle combinations: a high-temperature helium-cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to-hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable hydrogen production rates with the high-temperature helium-cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor. (authors)

  9. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  10. Accident analyses for china pressurizer reactor with an innovative conceptual design of passive residual heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Mingjun; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Qiu, Suizheng; Tian, Wenxi; Su, Guanghui

    2014-06-01

    Highlights: • An innovative passive system design concept was proposed. • Remote trigger control was adopted in the new passive system. • Transient characteristic of new PRHRS was studied during normal operation, SBO and FLB. • New design type is successful. - Abstract: An innovative passive safety system design concept for CPR 1000 nuclear power plant (NPP), proposed based on the original passive residual heat removal system (PRHRS) design, is presented and studied in this paper. The new type PRHRS avoids use of trigger valve inside the containment and adds an external water tank as the trip of PRHRS operation. The new design concept can realize functions of system remote control and trigger, while the external water tank can provide emergency water for steam generator (SG) in case of loss of feed water accident. Also, the advantages to use a valve and a tank located externally to containment are including the more reliability in implementing valve in pipes at low pressure, possibility operating manually the valve to start the system, refilling the tank in case of long time cooling and maintaining the valve easily. Best-estimate transient simulation code Relap5/MOD3.4 is applied to study the behavior of new design PRHRS and transient characteristics of primary loop system during normal condition and accident conditions. The transient processes of Station Black-Out (SBO) and Feed-water Line Break (FLB) accidents are studied to verify the function of new design PRHRS. Results show that the effect of valve position change from inside to outside containment can be ignored for the new design PRHRS in normal operation and the external valve can trip the operation of PRHRS in case of accidents. The new design PRHRS can remove core residual heat from primary loop effectively through establishing the stable natural circulations in the primary loop and PRHRS loop when the SBO and FLB accidents occur. It also can supply the emergency water to the SG in the event of

  11. Performance of a passive emergency heat removal system of advanced reactors in two-phase flow and with high concentration of non-condensable; Atuacao de um sistema passivo de remocao de calor de emergencia de reatores avancados em escoamento bifasico e com alta concentracao de nao-condensaveis

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Luiz Alberto

    2008-07-01

    The research and the development of passive emergency cooling systems are necessary for the new generation of thermo-nuclear systems. Some basic information on the operation of these systems require the research of some relative processes to the natural circulation, mainly in conditions of two-phase flow involving processes of condensation in the presence of non-condensable gases, because many found situations are new. The experimental facility called Bancada de Circulacao Natural (BCN) was used for the realization of tests with diverse concentrations of non-condensable and power levels. The non-condensable gas present in the circuit decreases the rate of heat transfer for the secondary of the heat exchanger, determining low efficiency of the heat exchanger. High concentration of non-condensable in the vapor condensation, determines negative pressure, and cause the inversion of the flow in the circuit. The initial concentration of non-condensable and the geometry of the circuit, in the inlet of the heat exchanger, determines the establishment of transitory with two-phase flow. The BCN was performed with the computational code of Analysis of Accidents and Thermal-Hydraulics RELAP5/MOD 3.3 and, the calculated values had been compared with the experimental data, presenting good agreement for small non-condensable concentrations. The values calculated for high concentrations of non-condensable had been satisfactory after the circuit to have reached the temperature of saturation in the electric heater. (author)

  12. Needs of nuclear data for advanced light water reactor

    International Nuclear Information System (INIS)

    Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores. (author)

  13. The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MITAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.

  14. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  15. a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

    Science.gov (United States)

    Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

    2009-08-01

    This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

  16. Advanced Fast Reactor - 100 (AFR-100) Report for the Technical Review Panel

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, Anton [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, Taek K. [Argonne National Lab. (ANL), Argonne, IL (United States); Middleton, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-06-04

    This report is written to provide an overview of the Advanced Fast Reactor-100 in the requested format for a DOE technical review panel. This report was prepared with information that is responsive to the DOE Request for Information, DE-SOL-0003674 Advanced Reactor Concepts, dated February 27, 2012 from DOE’s Office of Nuclear Energy, Office of Nuclear Reactor Technologies. The document consists of two main sections. The first section is a summary of the AFR-100 design including the innovations that are incorporated into the design. The second section contains a series of tables that respond to the various questions requested of the reactor design team from the subject DOE RFI.

  17. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

  18. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  19. Simulation for Supporting Scale-Up of a Fluidized Bed Reactor for Advanced Water Oxidation

    OpenAIRE

    Farhana Tisa; Abdul Aziz Abdul Raman; Wan Mohd Ashri Wan Daud

    2014-01-01

    Simulation of fluidized bed reactor (FBR) was accomplished for treating wastewater using Fenton reaction, which is an advanced oxidation process (AOP). The simulation was performed to determine characteristics of FBR performance, concentration profile of the contaminants, and various prominent hydrodynamic properties (e.g., Reynolds number, velocity, and pressure) in the reactor. Simulation was implemented for 2.8 L working volume using hydrodynamic correlations, continuous equation, and simp...

  20. Characterization of plasma-induced phenol advanced oxidation process in a DBD reactor

    OpenAIRE

    Marotta, E.; Ceriani, E.; Shapoval, V.; Schiorlin, M.; Ceretta, C.; Rea, M.; Paradisi, C.

    2011-01-01

    Abstract Using phenol as a model organic pollutant we studied and characterized an innovative advanced oxidation process in water using a prototype dielectric barrier discharge (DBD) reactor in which electrical discharges are produced in the air above the water surface. Phenol is decomposed quite efficiently in this reactor operated at room temperature and atmospheric pressure. The process selectivity to form CO2 is, however, to be improved since a large fraction of the treated org...

  1. Proceedings of the 1992 topical meeting on advances in reactor physics

    International Nuclear Information System (INIS)

    This document, Volume 1, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Code Benchmarks and Validation; Fuel Management; Nodal Methods for Diffusion Theory; Criticality Safety and Applications and Waste; Core Computational Systems; Nuclear Data; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual papers have been cataloged separately. (FI)

  2. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  3. The microbial community of a passive biochemical reactor treating arsenic, zinc and sulfate-rich seepage

    Directory of Open Access Journals (Sweden)

    Susan Anne Baldwin

    2015-03-01

    Full Text Available Sulfidogenic biochemical reactors for metal removal that use complex organic carbon have been shown to be effective in laboratory studies, but their performance in the field is highly variable. Successful operation depends on the types of microorganisms supported by the organic matrix, and factors affecting the community composition are unknown. A molecular survey of a field-based biochemical reactor that had been removing zinc and arsenic for over six years revealed that the microbial community was dominated by methanogens related to Methanocorpusculum sp. and Methanosarcina sp., which co-occurred with Bacteroidetes environmental groups, such as Vadin HA17, in places where the organic matter was more degraded. The metabolic potential for organic matter decomposition by Ruminococcaceae was prevalent in samples with more pyrolysable carbon. Rhodobium- and Hyphomicrobium-related genera within the Rhizobiales Order that have the metabolic potential for dark hydrogen fermentation and methylotrophy, and unclassified Comamonadaceae were the dominant Proteobacteria. The unclassified environmental group Sh765B-TzT-29 was an important Delta-Proteobacteria group in this BCR, that co-occurred with the dominant Rhizobiales OTUs. Organic matter degradation is one driver for shifting the microbial community composition and therefore possibly the performance of these bioreactors over time.

  4. Advanced Reactor Safety Program – Stakeholder Interaction and Feedback

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    In the Spring of 2013, the Idaho National Laboratory (INL) began discussions with industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling and simulation could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  5. Advanced reactor safety program. Stakeholder interaction and feedback

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H.; Smith, Curtis L.

    2014-08-01

    In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  6. Advanced fusion reactor design using remountable HTc SC magnet

    International Nuclear Information System (INIS)

    A new concept of fusion reactor design is proposed using remountable high critical temperature (HTc) superconducting (SC) magnet. There are two advantages using this system. First one is that the magnet system can be composed by parts, which means it easy to replace the damaged magnet module. The second one is that it becomes possible to access the reactor first wall easily. In order to realize this system, we have performed experiments using HTc SC tape. The experimental results indicate that the resistance of the jointed region becomes about 60 μΩ, which shows the feasibility of this concept. Using this system the remountable first wall system also has the feasibility based on thermomechanical analysis. (author)

  7. The US NRC Advanced Gas REactor Evaluator (AGREE)

    International Nuclear Information System (INIS)

    A Prismatic Module Reactor (PMR) core fluids model has been developed in support of the US NRC Next Generation Nuclear Plant (NGNP) evaluation model. The PMR core fluids modeling relies on a subchannel approach in which the primary coolant flowpath through the core region and vertical in-core and ex-core gaps can be modeled as individual subchannels. These subchannels are connected together to represent a three dimensional reactor. An initial validation calculation has been performed using data available in literature for bypass flow. The predicted bypass flow was within 2.6% of the value reported in the literature. Further verification and validation is awaiting the completion of the coupling of the fluids model with a triangular solid heat conduction model. (author)

  8. Advanced methods for nuclear reactor gas laser coupling

    Energy Technology Data Exchange (ETDEWEB)

    Miley, G.H.; Verdeyen, J.T.

    1978-06-01

    Research is described that led to the discovery of three nuclear-pumped lasers (NPLs) using mixtures of Ne--N/sub 2/, He--Hg, and He or Ne with CO or CO/sub 2/. The Ne--N/sub 2/ NPL was the first laser obtained with modest neutron fluxes from a TRIGA reactor (vs fast burst reactors used elsewhere in such work), the He--Hg NPL was the first visible nuclear-pumped laser, while the Ne--CO and He--CO/sub 2/ lasers are the first to provide energy storage on a millisecond time scale. Important potential applications of NPLs include coupling and power transmission from remote power stations such as nuclear plants in satellites and neutron-feedback operation of inertial confinement fusion plants.

  9. Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%

  10. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  11. STAR-C Space Thermionic Advanced Reactor - Compact

    International Nuclear Information System (INIS)

    STAR-C is a small, compact nuclear space power system that can be configured to provide 5 to 25 kW of electrical power for long duration space applications. A 10 kWe baseline design concept has been defined that has a 550 kg mass and is one meter long and 2/3 meter in diameter. The reactor configuration is based on the Soviet ''ROMASHKA'' reactor that was built and operated in the early 1960's. The thermionic power conversion as based on the U.S. Solar Energy Thermionic (SET) program. The reactor fuel is derived from the NERVA nuclear rocket program. A number of military spacecraft are under development for deployment in the 1990's. Power requirements for many of these systems range from 5 to 25 kWe. Typically, these platforms require low levels of power for continuous housekeeping functions and higher levels for alert and battle engagement conditions. It is highly desirable that the selected space power system have a substantial growth capability that can be utilized as the spacecraft concept matures. There is also an increased emphasis on requirements for system survivability to hostile weapons threats and the capability for spacecraft maneuverability for evasive action. This requires that the space power system must be compact and that it must be closely integrated into the spacecraft structure to avoid inertial and dynamic effects associated with the use of extendable structures and booms

  12. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    International Nuclear Information System (INIS)

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

  13. Advances toward a transportable antineutrino detector system for reactor monitoring and safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Reyna, D. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Lund, J.; Kiff, S.; Cabrera-Palmer, B. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bowden, N. S.; Dazeley, S.; Keefer, G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2011-07-01

    Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20-30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory - containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our below ground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino

  14. Progress in development and design aspects of advanced water cooled reactors

    International Nuclear Information System (INIS)

    The objective of the Technical Committee Meeting (TCM) was to provide an international forum for technical specialists to review and discuss technology developments and design work for advanced water cooled reactors, safety approaches and features of current water cooled reactors and to identify, understand and describe advanced features for safety and operational improvements. The TCM was attended by 92 participants representing 18 countries and two international organizations and included 40 presentations by authors of 14 countries and one international organization. A separate abstract was prepared for each of these presentations. Refs, figs, tabs

  15. Compiled reports on the applicability of selected codes and standards to advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

  16. A passively-safe fusion reactor blanket with helium coolant and steel structure

    International Nuclear Information System (INIS)

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ''beryllium-joint'' concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket

  17. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: 'contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).' This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  18. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  19. Advanced control of propylene polimerizations in slurry reactors

    Directory of Open Access Journals (Sweden)

    Bolsoni A.

    2000-01-01

    Full Text Available The objective of this work is to develop a strategy of nonlinear model predictive control for industrial slurry reactors of propylene polymerizations. The controlled variables are the melt index (polymer quality and the amount of unreacted monomer (productivity. The model used in the controller presents a linear dynamics and a nonlinear static gain given by a neuronal network MLP (multilayer perceptron. The simulated performance of the controller was evaluated for a typical propylene polymerization process. It is shown that the performance of the proposed control strategy is much better than the one obtained with the use of linear predictive controllers for setpoint tracking control problems.

  20. Status of advanced light water cooled reactor designs 1996

    International Nuclear Information System (INIS)

    The present report, which is significantly more comprehensive than the previously one, addresses the rationale and basic motivations that lead to a continuing development of nuclear technology, provides an overview of the world status of current LWRs, describes the present market situations, and identifies desired characteristics for future plants. The report also provides a detailed description of utility requirements that largely govern today's nuclear development efforts, the situation with regard to enhanced safety objectives, a country wise description of the development activities, and a technical description of the various reactor designs in a consistent format. The reactor designs are presented in two categories: (1) evolutionary concepts that are expected to be commercially available soon; and (2) innovative designs. The report addresses the main technical characteristics of each concept without assessing or evaluating them from a particular point of view (e.g. safety or economics). Additionally, the report identifies basic reference documents that can provide further information for detailed evaluations. The report closes with an outlook on future energy policy developments

  1. Advanced sodium fast reactor accident source terms : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  2. The microbial community of a passive biochemical reactor treating arsenic, zinc, and sulfate-rich seepage.

    Science.gov (United States)

    Baldwin, Susan Anne; Khoshnoodi, Maryam; Rezadehbashi, Maryam; Taupp, Marcus; Hallam, Steven; Mattes, Al; Sanei, Hamed

    2015-01-01

    Sulfidogenic biochemical reactors (BCRs) for metal removal that use complex organic carbon have been shown to be effective in laboratory studies, but their performance in the field is highly variable. Successful operation depends on the types of microorganisms supported by the organic matrix, and factors affecting the community composition are unknown. A molecular survey of a field-based BCR that had been removing zinc and arsenic for over 6 years revealed that the microbial community was dominated by methanogens related to Methanocorpusculum sp. and Methanosarcina sp., which co-occurred with Bacteroidetes environmental groups, such as Vadin HA17, in places where the organic matter was more degraded. The metabolic potential for organic matter decomposition by Ruminococcaceae was prevalent in samples with more pyrolyzable carbon. Rhodobium- and Hyphomicrobium-related genera within the Rhizobiales order that have the metabolic potential for dark hydrogen fermentation and methylotrophy, and unclassified Comamonadaceae were the dominant Proteobacteria. The unclassified environmental group Sh765B-TzT-29 was an important Delta-Proteobacteria group in this BCR that co-occurred with the dominant Rhizobiales operational taxonomic units. Organic matter degradation is one driver for shifting the microbial community composition and therefore possibly the performance of these bioreactors over time. PMID:25798439

  3. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

    2015-09-15

    Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

  4. CFD analysis of flow and temperature distribution inside the calandria of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Passive systems are being examined for the future AHWR reactor designs. One of these systems is the passive moderator cooling system, which removes heat from the moderator in case of a Station Black Out (SBO). The heavy-water moderator gets heated due to the residual heat from the core structures and rises upward due to buoyancy. This is cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator (Calandria) pressure beyond safe limits. In this paper CFD investigations are carried out to study the temperature distributions and flow distribution inside the Calandria using a three-dimensional CFD code, OpenFoam 2.2.0. The results provide a band of operable mass flow rates which are safe for operation by virtue of prediction of hot spots in the Calandria. (author)

  5. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  6. Experimental investigations of thermal-hydraulic processes arising during operation of the passive safety systems used in new projects of nuclear power plants equipped with VVER reactors

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.; Kalyakin, D. S.

    2014-05-01

    The results obtained from experimental investigations into thermal-hydraulic processes that take place during operation of the passive safety systems used in new-generation reactor plants constructed on the basis of VVER technology are presented. The experiments were carried out on the model rigs available at the Leipunskii Institute for Physics and Power Engineering. The processes through which interaction occurs between the opposite flows of saturated steam and cold water moving in the vertical steam line of the additional system for passively flooding the core from the second-stage hydro accumulators are studied. The specific features pertinent to undeveloped boiling of liquid on a single horizontal tube heated by steam and steam-gas mixture that is typical for of the condensing operating mode of a VVER reactor steam generator are investigated.

  7. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report

    International Nuclear Information System (INIS)

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

  8. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  9. Advanced Electroactive Single Crystal and Polymer Actuators for Passive Optics Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Large stroke and high precision electroactive single crystal and polymer actuators are desired for cryogenic passive optics such as Fabry-Perot Interferometer (FPI)...

  10. 78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

    Science.gov (United States)

    2013-01-28

    ... COMMISSION Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115... No. R-115, for the University of Illinois Advanced TRIGA Reactor (ATR). The NRC has terminated the..., Facility Operating License No. R-115 is terminated. The above referenced documents may be examined,...

  11. Passive heat removal characteristics of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jae Kwang; Kang, Hyung Seok; Yoon, Joo Hyun; Kim, Hwan Yeol; Cho, Bong Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRS of the SMART has excellent passive heat removal characteristics. 2 refs., 4 figs., 1 tab. (Author)

  12. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    The paper presents the current status of the Gas cooled Fast Reactor system development which is shared within the Generation IV International Forum including EURATOM through the 7th Framework Programme project GoFastR. The various areas considered will include suitable fuel compounds and high temperature resistant cladding materials options, core design optimisation, primary system boundary, energy conversion. The safety approach, mainly oriented on core cooling for the moment, will be recalled together with a discussion of the results obtained. Further potential improvements or simplification of the system safety, at the light of the Fukushima accident, including an indirect coupled cycle for the energy conversion and a self sustainable Decay Heat Removal loop will be mentioned. The main issues related to the necessary R&D programme accompanying the system development will be recalled (fuel and materials, helium coolant technology, components such as gas circulators, valves and heat exchangers, thermal barriers). (author)

  13. Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Sivill, R.L.

    1990-03-01

    This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

  14. A Dual-Stage Hydrothermal Flow Reactor for Green and Sustainable Synthesis of Advanced Hybrid Nanomaterials

    DEFF Research Database (Denmark)

    Hellstern, Henrik Christian Lund

    2016-01-01

    can be synthesized hydrothermally in a dual-stage flow reactor that is both environmentally benign and capable of producing high quantities which is a prerequisite for use in applications. A dual-stage hydrothermal flow reactor was developed for this purpose and used to produce hybrid nanomaterials...... of polycrystalline MoS2 nanosheets. The feasibility of producing supported nanomaterials is demonstrated using TiO2 nanocatalysts on MoS2 sheets This dissertation describes the dual-stage hydrothermal flow reactor and how advanced nanocomposites in high yields may be readily synthesized for potential use...... of differing morphologies: from smaller particles grafted on a larger particle for support, to spherical core-shell nanoparticles of 10-30 nm in diameter of a narrow size distribution. This is accomplished by synthesizing the core and shell in separate reactor zones to avoid a mixed product of ungrafted...

  15. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-02-22

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling... October 14, 2009 (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available...

  16. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-03-09

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling... October 14, 2009, (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available...

  17. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    International Nuclear Information System (INIS)

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs

  18. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-03-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

  19. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  20. 2015 Groundwater Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Ponds

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    This report summarizes radiological monitoring results from groundwater wells associated with the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds Reuse Permit (I-161-02). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  1. 2014 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Mike [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  2. Advanced and Innovative Reactor Concept Designs, Associated Objectives and Driving Forces

    International Nuclear Information System (INIS)

    Advanced and innovative options for fast reactors are presented through a short selection of recent publications at international conferences. Driving forces and major trends are analysed to give a comprehensive overview of the various existing projects and supportive R and D. (author)

  3. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Higgins, J.; Nasta, K.

    1997-01-01

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs.

  4. Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

    Science.gov (United States)

    Towell, Rusty; Niffte Collaboration

    2015-10-01

    The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

  5. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  6. The economics of advanced fuel cycles in CANDU (PHW) reactors

    International Nuclear Information System (INIS)

    The economic assessments of advanced fuel cycles performed within Ontario Hydro are collated and summarized. The results of the analyses are presented in a manner designed to provide a broad perspective of the economic issues regarding the advanced cycles. The enriched uranium fuel cycle is shown to be close to competitive at today's uranium prices, and its relative position vis-a-vis the natural uranium cycle will improve as uranium prices continue to rise. In the longer term, the plutonium-topped thorium cycle is identified as being the most economically desirable. It is suggested that this cycle may not be commercially attractive until the second or third decade of the next century. (auth)

  7. DOE/NE University Program in robotics for advanced reactors

    International Nuclear Information System (INIS)

    The US Department of Energy has provided support to four universities and the Oak Ridge National Laboratory in order to pursue research leading to the development and deployment of advanced robotic systems capable of performing tasks that are hazardous to humans, that generate significant occupational radiation exposure, and/or whose execution times can be reduced if performed by an automated system. The goal is to develop a generation of advanced robotic systems capable of economically performing surveillance, maintenance, and repair tasks in nuclear facilities and other hazardous environments. The approach to achieving the program objective is a transition from teleoperation to the capability of autonomous operation within three successive generations of robotic systems. The strategy adopted in order to achieve the program goals in an efficient and timely manner consists in utilizing, and advancing where required, state-of-the-art robotics technology through close interaction between the universities and the manufacturers and operators of nuclear power plants. There is a potentially broad range of applications for the robotic systems developed in the course of this project. Therefore, it is expected that efforts to obtain additional support from other agencies, e.g., DOD and NASA, will be successful. Areas of cooperation with other nations (e.g., Japan, France, Germany) are being explored. This Program features a unique teaming arrangement among the Universities of Florida, Michigan, Tennessee, Texas, and the Oak Ridge National Laboratory, and their industrial partners, Odetics, Gulf State Utilities, Florida Power and Light Company, Remotec, and Telerobotics International

  8. Status and some safety philosophies of the China advanced research reactor CARR

    Energy Technology Data Exchange (ETDEWEB)

    Luzheng Yuan [China Inst. of Atomic Energy, Beijing, BJ (China). Reactor Engineering Research and Design Dept.

    2001-07-01

    The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21{sup st} century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D{sub 2}O reflector not less than 8 x 10{sup 14} n/cm{sup 2}. s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10{sup 15} n/cm{sup 2}. s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

  9. Design and layout decisions for refuelling system of advanced fast neutron reactor

    International Nuclear Information System (INIS)

    The experience in operation of BOR-60, BN-350 and BN-600 power units, as well as development of refuelling systems for BN-800 power unit, allows developing of refuelling system for BN-1200 advanced reactor of new generation. The refuelling system was developed on the basis of possible technical decisions aimed at improvement of safety and technical-and-economic indices. Structural layout of BN-1200 reactor refuelling system is given. Main differences in BN-1200 reactor refuelling system as compared with BN-800 reactor are given. Design features of refuelling equipment are: - BN-1200 reactor has a split large rotating plug to allow transporting of its components by railway with subsequent assembling at site; - the refuelling box is fabricated in the form of sectional parallelepiped to allow transporting of its components by railway with subsequent assembling at site; - one 'direct' refuelling mechanism and one cantilever' refuelling mechanism are used to refuel rarely replaced protection assemblies that allows reducing of overall dimensions of rotating plugs; - the vertical elevator is arranged on the oval plug installed on the reactor cover. The upper structure with elevator drive rotates together with the elevator plug under rotary drive located on the oval plug. The vertical elevator allows sufficient reduction of refuelling box; - the refuelling machine runs on straight-line rails. The vertical elevator, gas gate valve on reactor refuelling channel, non-use of spent FA drum and enhanced radiation protection on the column of refuelling box machine allows reduction of specific materials consumption of BN-1200 reactor refuelling system by more than 10 times as compared with BN-800 reactor. To verify refuelling equipment operability the following experiments are planned: - mastering of gripper design for 'direct' refuelling mechanism and refuelling machine; - mastering of 'cantilever' for refuelling mechanism; - mastering of fresh FA conveyor design. As for the

  10. Technical Research on Safety Management and Effective Application of China Advanced Research Reactor

    International Nuclear Information System (INIS)

    China Advanced Research Reactor (CARR) is a tank in pool type, light water cooled, heavy water reflected research reactor. The maximum thermal neutron flux of the reactor is 1.0x1015 cm-2s-1, and the reactor power is 60 MW. The reactor was designed and constructed completely by China Institute of Atomic Energy (CIAE). The construction project began on Aug. 26, 2002, reactor criticality was achieved on May 13, 2010, and it is scheduled to complete power increasing tests by the end of 2011. Future operation of CARR is preparing and its utilization program is considered. It is expected that CARR will greatly improve and enhance the comprehensive research capability of nuclear science and technology and push the peaceful use of nuclear technology forward. The paper briefly presents the reactor safety features, the operation organization and responsibilities, the management of operation safety, and the future utilizations. According to national safety regulations of research reactor, evaluation of operation safety of CARR shall be executed after initial operation at power level and submit the revised ''Final Safety Analysis Report'' (FSAR) to the regulatory body.Ordinary operation shall be approved and operation license shall be issued by the regulatory body after review on the ''Final Safety Analysis Report.'' Vertical and horizontal channels with associated equipment and instruments are installed in reactor core and in heavy water reflector. CARR will be used to produce variety of RIs in comprehensive fields, to meet the requirements of engineering tests and irradiation for developing NPP fuels and materials in China, to apply for NTD of mono-crystalline silicone, NAA, neutron photography and to provide high intense neutron beam for application of neutron scattering experiments in an adequate scale and others, etc. (author)

  11. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  12. Field observation of advance warning/advisory signage for passive railway crossings with restricted lateral sightline visibility: an experimental investigation.

    Science.gov (United States)

    Ward, N J; Wilde, G J

    1995-04-01

    This study evaluated a newly proposed series of signs intended for passive crossings with restrictions to lateral sightline visibility. These signs provide advance warning of a crossing and the restriction to lateral visibility. In addition, the signs advise motorists to come to a complete stop before crossing. Motorist behaviour was examined before and after installation of these signs at a rural passive crossing. A second site was observed in parallel to control partially for any confounding effects. Results indicated that motorists reduced speed and searched approach quadrants longer at points in the approachway after installation of the signs. However, there was no reliable increase in the number of motorists coming to complete stop, engaging in search behaviours, or classified as safe. The results are discussed in terms of reasons for the lack of compliance with the sign advisory.

  13. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  14. An approach for assessing ALWR passive safety system reliability

    International Nuclear Information System (INIS)

    Many of the advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive rather than active systems to perform safety functions. Despite the reduced redundancy of the passive systems as compared to active systems in current plants, the assertion is that the overall safety of the plant is enhanced due to the much higher expected reliability of the passive systems. In order to investigate this assertion, a study is being conducted at Sandia National Laboratories to evaluate the reliability of ALWR passive safety features in the context of probabilistic risk assessment (PRA). The purpose of this paper is to provide a brief overview of the approach to this study. The quantification of passive system reliability is not as straightforward as for active systems, due to the lack of operating experience, and to the greater uncertainty in the governing physical phenomena. Thus, the adequacy of current methods for evaluating system reliability must be assessed, and alternatives proposed if necessary. For this study, the Westinghouse Advanced Passive 600 MWe reactor (AP600) was chosen as the advanced reactor for analysis, because of the availability of AP600 design information. This study compares the reliability of AP600 emergency cooling system with that of corresponding systems in a current generation reactor

  15. Conceptual designs of advanced fast reactor. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    A Technical Committee meeting (TCM) was held on Conceptual Designs of Advanced Fast Power Reactors to review the lessons learned from the construction and operation of demonstration and near-commercial size plants. This TCM focused on design and development of advanced fast reactors and identified methodologies to evaluate the economic competitiveness and reliability of advanced projects. The Member States which participated in the TCM were at different stages of LMFR development. The Russian Federation, Japan and India had prototype and/or experimental LMFRs and continue with mature R and D programmes. China, the Republic of Korea and Brazil were at the beginning of LMFR development. Therefore the aims of the TCM were to obtain technical descriptions of different design approaches for experimental, prototype, demonstration and commercial LMFRs, and to describe the engineering judgements made in developing the design approaches. Refs, figs, tabs

  16. MAAP-impair interface for analysis of iodine behavior in advanced reactor accidents

    International Nuclear Information System (INIS)

    As part of the US Department of Energy (US DOE) Advanced Reactor Severe Accident Program, a study was initiated to provide an ex-vessel iodine analytical capability to estimate source terms for severe accidents in advanced light water reactors. This capability has been developed by creating a software link, MID, between the MAAP and IMPAIR computer codes. The interface allows IMPAIR to access the thermal-hydraulic and fission product results provided by MAAP and use these results to drive the chemical reaction and physical mass transfer models in IMPAIR. The first phase of the development is designed to provide iodine analytical capability up to the point of reactor vessel failure. A follow-on study is planned to address iodine behavior in accident scenarios that go beyond vessel failure. A number of MAAP-IMPAIR demonstration calculations have been performed for the General Electric simplified boiling water reactor and Westinghouse AP600 reactor designs. These calculations demonstrated that the software interface provided the necessary link to create a functional ex-vessel iodine analytic capability. They also clearly indicated that both the chemical and the physical behavior of iodine species in the containment are strongly dependent upon the containment thermal-hydraulic conditions

  17. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  18. Advanced Fusion Reactors for Space Propulsion and Power Systems

    Science.gov (United States)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  19. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, FBRs are to be developed as the main of future nuclear power generation in Japan, and when the development is advanced, it is positivity aimed at building up the plutonium utilization system using FBRs superior to the uranium utilization system with LWRs. Also it was decided that it is necessary to exert incessant effort for the development of FBRs under the proper cooperation system of the government and people for a considerable long period, and as for the concrete development, hereafter, the deliberation is advanced by the expert subcommittee on FBR development project of the Atomic Energy Commission in succession. The subcommittee was founded in May, 1986, to carry out the deliberation on the long term promotion measures for the development of FBRs, the promotion measures for the research and development, the evaluation and examination of the basic specification of a demonstration FBR, the promotion measures for the international cooperation and other important matters related to the development of FBRs. The construction of the prototype FBR 'Monju' is in progress aiming at the criticality in 1992, and the start of construction of a demonstration FBR is expected in the latter half of 1990s. The situation around the development of FBRs, the fundamentals for promoting the research and development, and the subjects of the research and development are reported. (Kako, I.)

  20. Advanced calculational methods for power reactors and LWR core design parameters

    International Nuclear Information System (INIS)

    The purpose of the Specialists Meeting on Advanced Calculational Methods for Power Reactors, held in Cadarache, France, 10-14 September 1990, was to provide a forum for reviewing and discussing selected core physics of water cooled reactors (including high convertors). New methods of advanced calculation for advanced fuels and complex geometries of next generation reactors with a high level of accuracy were discussed and the importance of supercomputing and on-line monitoring was also acknowledged. The meeting was attended by about 60 participants from 20 countries who presented 30 papers. The Technical Committee Meeting on LWR Core Design Parameters, held in Rez, former Czechoslovakia, 7-11 October 1991, provided an opportunity for participants to exchange their experience on reactor physics aspects of benchmark calculations of various lattices, methods for core parameter calculations, core monitoring and in-core fuel management. At the Workshop there were further discussions related to the benchmark problems, homogenization techniques and cross-section representations. Thirty-five papers were presented by about 43 participants from 19 countries. A separate abstract was prepared for each of the mentioned papers. Refs, figs and tabs

  1. Advanced Passivation Technology and Loss Factor Minimization for High Efficiency Solar Cells.

    Science.gov (United States)

    Park, Cheolmin; Balaji, Nagarajan; Jung, Sungwook; Choi, Jaewoo; Ju, Minkyu; Lee, Seunghwan; Kim, Jungmo; Bong, Sungjae; Chung, Sungyoun; Lee, Youn-Jung; Yi, Junsin

    2015-10-01

    High-efficiency Si solar cells have attracted great attention from researchers, scientists, photovoltaic (PV) industry engineers for the past few decades. With thin wafers, surface passivation becomes necessary to increase the solar cells efficiency by overcoming several induced effects due to associated crystal defects and impurities of c-Si. This paper discusses suitable passivation schemes and optimization techniques to achieve high efficiency at low cost. SiNx film was optimized with higher transmittance and reduced recombination for using as an effective antireflection and passivation layer to attain higher solar cell efficiencies. The higher band gap increased the transmittance with reduced defect states that persisted at 1.68 and 1.80 eV in SiNx films. The thermal stability of SiN (Si-rich)/SiN (N-rich) stacks was also studied. Si-rich SiN with a refractive index of 2.7 was used as a passivation layer and N-rich SiN with a refractive index of 2.1 was used for thermal stability. An implied Voc of 720 mV with a stable lifetime of 1.5 ms was obtained for the stack layer after firing. Si-N and Si-H bonding concentration was analyzed by FTIR for the correlation of thermally stable passivation mechanism. The passivation property of spin coated Al2O3 films was also investigated. An effective surface recombination velocity of 55 cm/s with a high density of negative fixed charges (Qf) on the order of 9 x 10(11) cm(-2) was detected in Al2O3 films. PMID:26726397

  2. Conceptual Design of Ultra-long-life Core Fast Reactor (UCFR) with Ga-based Passive Decay Heat Removal System (PDHRS)

    International Nuclear Information System (INIS)

    UCFR (Ultra-long-life Core Fast Reactor) is a 260MWth /100MWe sodium-cooled fast reactor which requires no on-site refueling and meets the need for future nuclear energy systems. UCFR is a pool type reactor with metallic fuels, four intermediate heat exchangers, two steam generators, and passive decay heat removal systems. Because gallium has the chemical reaction safety such as low oxygen reactivity compared to sodium, it can be used as a boundary material between sodium and atmosphere to enhance the nuclear safety of UCFR. In this research, design studied for neutronics and thermal-hydraulics are included. The safety performance of UCFR will be analyzed with MARS-LMR. Although MARS-LMR was originally intended for a safety analysis of liquid metal-cooled reactor, gallium properties were newly added to this code which is applicable for gallium-cooled systems. The properties of various liquid metals are indicated in table II. Considering needs to improve uranium utilization and solve the nuclear proliferation, ultra-long cycle fast reactor has been developed. UCFR is a 260MWth/100MWe sodium-cooled fast reactor which requires no on-site refueling during design period with metallic fuels (U-5Zr), HT-9 cladding, four intermediate heat exchangers, two steam generators, and Ga-based PDHRS. Through this paper, new PDHRS using gallium that can be remove decay heat passively for an infinite time is suggested. In Ga-based PDHRS, the both water and air as an ultimate heat sink will be can be considered because gallium has the chemical reaction safety. In this research, design study for neutronics and thermal-hydraulics were also included. For safety analysis of UCFR using MARS-LMR, detailed design of UCFR with Ga-based PDHRS will be required

  3. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  4. Engineering, safety, and economic evaluations of ASPIRE [Advanced Safe Pool Immersed REactor

    International Nuclear Information System (INIS)

    A preconceptual design of a tokamak fusion reactor concept called ASPIRE (Advanced Safe Pool Immersed REactor) has been developed. This concept provides many of the attractive features that are needed to enhance the capability of fusion to become the power generation technology for the 21st century. Specifically, these features are: inherent safety, low pressure, environmental compatibility, moderate unit size, high availability, high thermal efficiency, simplicity, low radioactive inventory, Class C radioactive waste disposal, and low cost of electricity. We have based ASPIRE on a second stability tokamak. However, the concept is equally applicable to a first stability tokamak or to most other magnetic fusion systems

  5. Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

    International Nuclear Information System (INIS)

    The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.)

  6. Fiscal year 1998 multi-year work plan. Advanced reactors transition program

    International Nuclear Information System (INIS)

    The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building

  7. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

    The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

  8. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  9. Overview of welding of oxide dispersion strengthened (ODS) alloys for advanced nuclear reactor applications

    International Nuclear Information System (INIS)

    Oxide dispersion strengthened (ODS) alloys are very promising materials for Generation IV reactors with a potential to be used at elevated temperatures under severe neutron exposure environment. Welding of the ODS alloys is an understudied problem. In this paper, an overview of welding of the ODS alloys useful for advanced nuclear reactor applications is presented. The microstructural changes and the resultant mechanical properties obtained by various solid state welding processes are reviewed. Based on our results on PM2000, an approach for future work on welding of the ODS alloys is suggested. (author)

  10. Preliminary design of an advanced programmable digital filter network for large passive acoustic ASW systems. [Parallel processor

    Energy Technology Data Exchange (ETDEWEB)

    McWilliams, T.; Widdoes, Jr., L. C.; Wood, L.

    1976-09-30

    The design of an extremely high performance programmable digital filter of novel architecture, the LLL Programmable Digital Filter, is described. The digital filter is a high-performance multiprocessor having general purpose applicability and high programmability; it is extremely cost effective either in a uniprocessor or a multiprocessor configuration. The architecture and instruction set of the individual processor was optimized with regard to the multiple processor configuration. The optimal structure of a parallel processing system was determined for addressing the specific Navy application centering on the advanced digital filtering of passive acoustic ASW data of the type obtained from the SOSUS net. 148 figures. (RWR)

  11. Advanced Demonstration of Motion Correction for Ship-to-Ship Passive Inspections

    Energy Technology Data Exchange (ETDEWEB)

    Ziock, Klaus-Peter [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Boehnen, Chris Bensing [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ernst, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2013-09-30

    Passive radiation detection is a key tool for detecting illicit nuclear materials. In maritime applications it is most effective against small vessels where attenuation is of less concern. Passive imaging provides: discrimination between localized (threat) and distributed (non-threat) sources, removal of background fluctuations due to nearby shorelines and structures, source localization to an individual craft in crowded waters, and background subtracted spectra. Unfortunately, imaging methods cannot be easily applied in ship-to-ship inspections because relative motion of the vessels blurs the results over many pixels, significantly reducing sensitivity. This is particularly true for the smaller water craft where passive inspections are most valuable. In this project we performed tests and improved the performance of an instrument (developed earlier under, “Motion Correction for Ship-to-Ship Passive Inspections”) that uses automated tracking of a target vessel in visible-light images to generate a 3D radiation map of the target vessel from data obtained using a gamma-ray imager.

  12. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  13. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  14. Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide

    International Nuclear Information System (INIS)

    In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done

  15. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  16. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

  17. Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

    2006-10-01

    Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

  18. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  19. Large-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    International Nuclear Information System (INIS)

    The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission's preapplication review of the PIUS reactor. Baseline calculations of the PIUS Supplement design were performed for a large-break loss-of-coolant (LBLOCA) initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following an LBLOCA

  20. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  1. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    International Nuclear Information System (INIS)

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  2. Effects of Plant Capital Costs on Market Introduction of Advanced Reactors

    International Nuclear Information System (INIS)

    In order to make reasonable forecasts about the future market shares of the different reactor types in the energy market, scenario studies are used. For this purpose, the future nuclear reactor park mix in Europe has been analysed applying an integrated dynamic process modelling technique. Starting point in the analyses is the current nuclear reactor park in the EU27 countries, taken into account the foreseen lifetime of each individual reactor as well as the nuclear fuel cycle infrastructure. Furthermore, an energy demand scenario, which is derived from the World Economic Council and the European DG-TREN, has been applied as input for the analyses. Various market share scenarios for nuclear energy are derived including sub-variants with regard to the intra-nuclear options taken, e.g. introduction date of Gen-III (i.e. EPR) and Gen-IV (i.e. SCWR, HTR, FR) reactors, level of reprocessing, and so forth. Realistic assumptions with respect to introduction dates of advanced reactors, reactor characteristics and fuel cycle facility characteristics were taken attempt-ing to provide an as realistic as possible framework for the assessment of future nuclear energy system scenarios in Europe. The assessment was undertaken using the DANESS code (Dynamic Analysis of Nuclear Energy System Strategies, developed by Argonne National Laboratory) and providing a complete picture of mass-flow and economics of the various nuclear energy system scenarios. The present assessment recognizes the integrated nuclear fuel cycle and foresees the evolution in cost of electricity. For the evaluation of the different future development paths for nuclear energy, reactor, front- and back- end specific cost factors are associated with the mass flow of uranium, plutonium, and minor actinides. The energy costing model covers capital cost, O and M costs and fuel cycle costs and consequently calculates levelized cost of electricity for the given nuclear system. The analyses show that the future

  3. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    Science.gov (United States)

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  4. Light Water Reactor Sustainability Program Advanced Seismic Soil Structure Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Bolisetti, Chandrakanth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    of interest. The specific nonlinear soil behavior included in the NLSSI calculation presented in this report is gapping and sliding. Other NLSSI effects are not included in the calculation. The results presented in this report document initial model runs in the linear and nonlinear analysis process. Final comparisons between traditional and advanced SPRA will be presented in the September 30th deliverable.

  5. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  6. Steam condensation model onto horizontal finned tubes: first approximation to the containment cooling system of advanced reactors European Designs

    International Nuclear Information System (INIS)

    European designs of advanced reactors, such as EPR pr SWR 1000, have considered the use of innovative passive safety systems to preserve containment integrity even in the case of a hypothetical accident. These systems consist of several units of bundles of quasi-horizontal finned tubes. Steam released into the containment atmosphere condenses onto these structures, which are internally cooled by water under natural circulation regime. The energy absorbed by the coolant is then discharged into a pool which acts as a heat sink for at least three days. This paper presents the work carried out under the auspices of European Union within the CONGA project to simulate steam condensation onto the above mentioned quasi-horizontal finned tubes. To date calculation methodologies have been pearly reviewed and and an approximation (''Nusselt type'') has been accepted to be the most suitable for safety studies, because of its mechanistic nature and its compatibility with current safety computation tools. Two versions of this approach have been properly adapted and subsequently implemented into independent codes for their validation. An experimental database built up from the open literature allowed to point out models accuracy, showing error well within the experimental uncertainly margin. Therefore, condensate film resistance to heat transfer has been modelled satisfactorily. Nevertheless, further work remains to be done to account for the effects of noncondensable gas presence and aerosol deposition onto heat transfer surfaces. (Author) 22 refs

  7. Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kiselyov, V.A.; Sokov, L.M.

    1997-04-01

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.

  8. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  9. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  10. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

  11. Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William Richins; Stephen Novascone; Cheryl O' Brien

    2009-08-01

    Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

  12. Design features facilitating the decommissioning of advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    The design of the advanced gas-cooled reactors is discussed as is the proposed decommissioning plan for delayed decommissioning. The special features which assist in decommissioning are presented. As a result of the study a catalogue of design features which will facilitate decommissioning is given. In addition to the catalogue of design features, the radioactive inventory 10 years after shutdown and 100 years after shutdown has been calculated. From this a provisional operator dose from activities associated with decommissioning has been assessed

  13. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, ''Comprehensive Emergency Management System.''

  14. The importance of collaboration in the advancement of current and next generation reactors

    International Nuclear Information System (INIS)

    The sections of the contribution are as follows: Tradition of innovation. Growing demand for nuclear power; Collaboration drivers; Responses. Knowledge transfer and management is critical. What kind of focus? Equipment reliability. Advanced repair, replacement and construction approaches. Materials. Plant safety margins. Spent fuel management. Examples of European collaboration. Zorita materials examination. Collaboration in the development of next generation reactors; Westinghouse R and D priorities; A look to the future. (P.A.)

  15. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE-ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  16. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  17. Reliability prediction for the vehicles equipped with advanced driver assistance systems (ADAS and passive safety systems (PSS

    Directory of Open Access Journals (Sweden)

    Balbir S. Dhillon

    2012-10-01

    Full Text Available The human error has been reported as a major root cause in road accidents in today’s world. The human as a driver in road vehicles composed of human, mechanical and electrical components is constantly exposed to changing surroundings (e.g., road conditions, environmentwhich deteriorate the driver’s capacities leading to a potential accident. The auto industries and transportation authorities have realized that similar to other complex and safety sensitive transportation systems, the road vehicles need to rely on both advanced technologies (i.e., Advanced Driver Assistance Systems (ADAS and Passive Safety Systems (PSS (e.g.,, seatbelts, airbags in order to mitigate the risk of accidents and casualties. In this study, the advantages and disadvantages of ADAS as active safety systems as well as passive safety systems in road vehicles have been discussed. Also, this study proposes models that analyze the interactions between human as a driver and ADAS Warning and Crash Avoidance Systems and PSS in the design of vehicles. Thereafter, the mathematical models have been developed to make reliability prediction at any given time on the road transportation for vehicles equipped with ADAS and PSS. Finally, the implications of this study in the improvement of vehicle designs and prevention of casualties are discussed.

  18. Advanced phase change materials and systems for solar passive heating and cooling of residential buildings

    Energy Technology Data Exchange (ETDEWEB)

    Salyer, I.O.; Sircar, A.K.; Dantiki, S.

    1988-01-01

    During the last three years under the sponsorship of the DOE Solar Passive Division, the University of Dayton Research Institute (UDRI) has investigated four phase change material (PCM) systems for utility in thermal energy storage for solar passive heating and cooling applications. From this research on the basis of cost, performance, containment, and environmental acceptability, we have selected as our current and most promising series of candidate phase change materials, C-15 to C-24 linear crystalline alkyl hydrocarbons. The major part of the research during this contract period was directed toward the following three objectives. Find, test, and develop low-cost effective phase change materials (PCM) that melt and freeze sharply in the comfort temperature range of 73--77{degree}F for use in solar passive heating and cooling of buildings. Define practical materials and processes for fire retarding plasterboard/PCM building products. Develop cost-effective methods for incorporating PCM into building construction materials (concrete, plasterboard, etc.) which will lead to the commercial manufacture and sale of PCM-containing products resulting in significant energy conservation.

  19. Summary of SMIRT20 Preconference Topical Workshop - Identifying Structural Issues in Advanced Reactors

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; (1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and (2) calibrating simulation software and methods that address topic 1. The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

  20. Trends in Advanced Reactor Development and the Role of the IAEA

    International Nuclear Information System (INIS)

    Energy supply is an important prerequisite for further socio-economic development, especially in developing countries where the per capita energy use is only a very small fraction of that in industrialized countries. Nuclear energy is an essentially unlimited energy resource with the potential to provide this energy in the form of electricity, district heat and process heat under environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power reactors. The experience forms a sound basis for further improvements. Nuclear programmes in many countries are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety in order to overcome the current concerns of nuclear power. Advanced reactors now being developed could help to meet the demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes

  1. Behaviour analysis of AC-600 passive safety systems

    International Nuclear Information System (INIS)

    Southwest Center of Reactor Engineering Research and Design has finished the first step conceptual design of 600 mwe advanced PWR (AC-600). The main research emphases of AC-600 conceptual design include the advanced reactor core, the passive safety systems and the simplification. The passive safety systems of AC-600 consist of two reactor make up water tanks, two accumulators, two emergency feedwater tanks, two emergency natural draft air condensers, a containment water jacket and an enhanced primary cycle natural circulation flow system. 25% of the rated reactor power can be removed by the natural circulation cooling. The full pressure reactor make up water tanks are able to provide enough borated water which would be injected into the reactor coolant system during small LOCA. The coolant natural circulations can be established in the primary system and the passive secondary emergency feedwater system, removing residual heat from the reactor core to the atmosphere when station blackout occurs. It is indicated from analysis that the containment diameter of AC-600 is about 35 m. The large tanks and the large vertical distances between the tanks and reactor core are the main reason of using the big containment. It is also indicated from analysis that the low head safety injection pumps are required in AC-600 design to assure the recirculation system operation when large LOCA occurs. The reliability of AC-600 engineered safety systems is increased because the function of the passive safety systems is conducted through the immutable natural laws. The paper discusses the natural circulation ability and safety behavior of the passive safety systems during LOCA or station blackout for AC-600. The passive limits to excess reactivity and thermal hydraulic transients are also preliminarily discussed. Figs and tabs

  2. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  3. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  4. The scaling of economic and performance parameters of DT and advanced fuel fusion reactors

    International Nuclear Information System (INIS)

    In this study, the plasma stability index beta and the fusion power density in the plasma were treated as independent variables to determine how they influenced three economic performance parameters of fusion reactors burning the DT and four advanced fusion fuel cycles. The economic/performance parameters included the total power produced per unit length of reactor; the mass per unit length, and the specific mass in kilograms/kilowatt. The scaling of these parameters with beta and fusion power density was examined for a common set of engineering assumptions on the allowable wall loading limits, the maximum magnetic field existing in the plasma, average blanket mass density, etc. It was found that the power per unit length decreased as the plasma power density and beta increased. This is a consequence of the fact that the first wall is a bottleneck in the energy flow from the plasma to the generating equipment, and the wall power flux will exceed wall loading limits if the plasma radius exceeds a critical value. If one wished to build an engineering test reactor which produced a burning plasma at the lowest possible initial cost, and without regard to whether such a reactor would ultimately produce the cheapest power, then one would minimize the mass per unit length. The mass per unit length decreases with increasing plasma power density and beta, with the DT reaction being the most expensive at a fixed plasma power density (because of its thicker blanket), and the least expensive at a fixed value of beta, at least up to values of beta of 50%. The specific mass, in kg/kw, which is a rough measure of the cost of the power generated by the reactor, shows an opposite trend. It increases with increasing plasma power density and beta. At a given plasma power density and low beta, the DT reaction gives the lowest specific mass, but at a fixed beta above 10%, the advanced fuel cycles have the lowest specific mass

  5. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  6. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    International Nuclear Information System (INIS)

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs

  7. The nuclear fuel cycle with advanced reactor systems - analysis of its economic fundamentals and possibilities

    International Nuclear Information System (INIS)

    The purpose of the study is to analyse the nuclear fuel cycle of alternative advanced reactor systems with respect to their different mass flows of nuclear fuel and to judge the economic feasibility of these advanced nuclear technologies using a specific fuel cycle model. It is the particular importance of this subject that many technical, physical, political and economic coherences are combined in a very complex manner. A detailed description of the problem is given in the introductional chapter 1. The following chapter 2 gives a sufficient survey of the different techniques and technical facilities of the nuclear fuel cycles in question. Part 3 includes an investigation of logical coherences between typical fuel cycle mass flows which consequently leads to a mathematical model. This model is described in part 4. Chapter 5 then deals with the application of this model by the quantitative estimation and valuation of the economic differences between the conventional and advanced nuclear technology. In the final part of this study the influence of a very important parameter in this context, the price of plutonium, is discussed with respect to the time of introduction of the advanced reactor technology under economic conditions. (orig.)

  8. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    International Nuclear Information System (INIS)

    Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

  9. Influence of Local Natural Circulation on Passive Residual Heat Removal in Nuclear Reactor%反应堆内局部自然循环对非能动余热排出的影响

    Institute of Scientific and Technical Information of China (English)

    于雷; 鄢炳火; 刘守相

    2012-01-01

    为研究反应堆堆内局部自然循环对非能动余热排出的影响,利用改进的RELAP5/MOD3.2程序对核动力装置及非能动余热排出系统进行数学建模与理论研究,并利用试验数据进行了校核.研究表明:在核动力装置自然循环运行条件下,由于反应堆上封头旁流及反应堆入口漏流通道的存在,在反应堆活性区、上封头、环腔及下腔室之间构成了局部自然循环流动现象;在主回路自然循环能力较弱时,堆内产生的局部自然循环流动占优,反应堆衰变热无法顺利带出.%The nuclear power plant and passive residual heat removal system were modeled and investigated with the advanced code developed from RELAP5/MOD3. 2 code, for analyzing the influence of local natural circulation on the passive residual heat removal in nuclear reactor. The simulation results were validated with experimental data. The results indicate that the local natural circulation occurs in the connected region consists of reactor radiation zone, upper head, annular channel and lower plenum in case of natural circulation operation, due to the upper head by-passing flow and the leak flow at the core inlet. As the natural circulation capability is weak, the local natural circulation in the core is so significant that the decay heat can not be removed successfully.

  10. Implications of Results from the Advanced Gas Reactor Fuel Development and Qualification Program on Licensing of Modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2001-10-01

    The high level of safety of modular high temperature gas-cooled reactor (HTGR) designs is achieved by passively maintaining core temperatures below fission product release thresholds under all envisioned accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to other reactor types but is predicated on exceptionally high coated-particle fuel fabrication quality and excellent fuel performance under normal operation and accident conditions. The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program decided to qualify for uranium oxide/uranium carbide (UCO) TRISO coated-particle fuel in an operating envelope that would bound both pebble bed and prismatic modular HTGR options. By using a mixture of uranium oxide and uranium carbide, the kernel composition is engineered to minimize CO formation and fuel kernel migration, which is key to maintain to fuel integrity at the higher burnups, temperatures, and temperature gradients anticipated in prismatic HTGRs. Fuel fabrication conducted at both laboratory and engineering scale has demonstrated the ability to fabricate high quality UCO TRISO fuel with very low defects. The first irradiation (AGR 1) exposed about 300,000 TRISO fuel particles to a peak burnup of 19.6% FIMA, a peak fast-neutron fluence of about 4.3 × 1025 n/m2, and a maximum time-averaged fuel temperature of about 1,200°C without a single particle failure. The very low release of key metallic fission products (except silver) measured in post-irradiation examination (PIE) confirms the excellent performance measured under irradiation. Very low releases have been measured in accident simulation heatup testing (''safety testing'') after hundreds of hours at 1600 and 1700°C and no particle failures (no noble gas release measured) have been observed. Even after hundreds of hours at 1800°C, the releases are

  11. A study of the advancement of a reactor core design environment

    International Nuclear Information System (INIS)

    Full text: During the years from 2002 to 2004 a joint project has been performed by IFE, Halden and Yonden Engineering Corporation, Japan, to develop an advanced reactor core design environment based on a communication method for controlling a reactor core code system efficiently from PCs in a distributed network. The advanced reactor core design environment is realized by using Microsoft Visual Basic and communication software based on the IFE product SoftwareBus. The project has been carried out based on the fact that a computer-aided design system has been under development at Yonden Engineering Corporation in order to perform efficiently fuel replacement calculation by Yonden's reactor design code system. In this system, the structure is such that the physics calculation code system runs on UNIX workstations (in parallel) performing the calculations, while the Man-Machine Interface for controlling the calculation programs run on PCs in a distributed network. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general communication tool (IFE's SoftwareBus) has been used for realizing communication of the n-pair n-node between the reactor core design code system and the PC applications. Further, a method of improvement in the speed of the optimal pattern calculation has been implemented by assigning each examination pattern to two or more computers distributed in the network and assigning the next pattern calculation to the computer, where the calculation has ended or has the lowest workload. The high-speed technology of the pattern survey by network distributed processing is based on SoftwareBus. The reactor core design code system is developed in FORTRAN running on a UNIX workstation (Solaris). The PC applications have been developed by using Microsoft Visual Basic on Windows 2000 platform. The first step of the verification and validation process was carried out in March

  12. Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

    2011-05-31

    Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus

  13. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  14. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the